Text

Forwards Util Proposed Changes to NRC Second Draft Tech Specs as Part of Ongoing Tech Spec Development Effort,Per 881019 Meeting W/Nrc
	ML20206G039
Person / Time
Site:	Comanche Peak
Issue date:	11/16/1988
From:	Beck J; TEXAS UTILITIES ELECTRIC CO. (TU ELECTRIC)
To:	; NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM)
References
	TXX-88780, NUDOCS 8811220012
	Download: ML20206G039 (124)
	v • d • e

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M M Log # TXX-88780

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.~ile # 10014

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916 clo Ref I 10CFR50.36 illELECTRIC

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November 16, 1988 Wiluem G. Cour"..

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U. S. Nuclear Regul6 tory Commission i

Attn: Document Control Desk Washington, D.C.

20555

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SUBJECT:

COMANCHE PEAK STEAM ELECTRIC STATION (CPSES) i DOCKET N0. 50-445 PROPOSED CHANGES TO THE NRC SECOND DRAFT TECHNICAL SPECIFICATIONS REF:

1)

C. 1. Grimes letter to W. G. Counsil dated March 2, 1988 2)

W. G. Counsil Ictter (TXX-88512) to NRC dated June 23, 1988 L

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Gentlemen:

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Reference 1 provided CPSES with a copy of the NRC's second draft Technical

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Specifications.

TU Electric reviewed this document and proposed changes in i

Reference 2.

Subsequent meetings were held with the NRC to discuss these E

changes. As part of the ongoing Technical Specification development efforts, I

and as discussed in a October 19, 1988 meeting with the NRC, attached are i

additional changes to the NRC's second draft Technical Specifications.

The l

attached pages are intended to supersede selected pages from our previous l

j submittal (Reference 2),

j Very truly yours, h

i W. G. Counsil By:

EM.

T/n W. Beck Vice President,

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Nuclear Engineering j

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Attachment c. Mr. R. D. Martin, Region IV (1 copy))

Resident inspectors, USES (3 copies Mr. R. F. Warnick, MC (1 cop )

[pp.i

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Mr. J. H. Wilson, NRC (1 copy

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Mr. Bob Giartiina, NRC (1 copy

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3 PDC I

h CPSES Technical Spocifications Pm2#0 NRC Orafe 2 Markup Section 1

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Change I0f Justification For Change Change #0001.

RDebtMeference to table 3.6-1 with the Technical spuuted b nx4780 Specification Diiprovement Program (TSIP) since the table 3.6-1 su "Mumf ry b.

All equipment hatches are closed and sealed, it-;. 000' I

c.

Each air lock is in compliance with the requirements of Specificat.i::n 3.6.1.3, d.

The containment leakage rates are within the limits of Specification 3.6.1.2, and e.

The sealing mechanism associated with each penetration (e.g., welds, bellows, or 0-rings) is OPERABLE.

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CONTROLLED LEAKAGE

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1.8 CONTROLLED LEAKAGE shall be that seal water flow supplied to the reactor coolant pump seals.

CORE ALTERATIONS 1.9 CORE ALTERATIONS shall be the movement or manipulation of any coeponent within the reactor pressure vessel with the vessel head removed an:) fuel in the vessel.

Suspension of CORE ALTERATIONS shall not preclude con lletion of movement of a component to a safe conservative position.

DIGITAL CAANNEL OPERATIONAL TEST 4

1.10 A DIGITAL CHANNEL OPERATIONAL TEST shall consist of exercising the digital computer hardware using data base manipulation and injecting simulate; process data to verify OPERABILITY of alarm and/or trip functions.

OOSE EQUIVALENT I-131 1.11 00SE EQUIVALENT I-131 shall be that concentration of I-131 (microcurie / gram) which alone would produce the same thyroid dose as the quantity and isotopic mixture of I-131,1-132,1-133, I 134, and I-135 actually present.

The thyroid dose conversion factors used for this calculation shall be those listed in Table III of TID-14844, "Calculation of Distance Factors for Power and Test Reactor Sites" or Table E-7 of NRC Regulatory Cuide 1.109. Revision 1, October 1977.

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CCMANCHC PEAK - UNIT 1 1-2

T u.-6t733 CPSES Technical Specifications RecumMT 2.

NRC Draft 2 Markup PRC4 2A oF to Section 2.0 Change ID#

Justification for Change 5102 Table 2.2-1 notations have been updated to account for the 5103 effects of gama streaming and steam generator modifications, t

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TABLE 2.2-1 (Continued) 2E= 1 n

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TABLE NOIATIONS (Continued) 4

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(Continued)

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P
=
Pressurizer pressure, psig, Il 7
P8 2235 psig (Nominal RCS operating pressure),
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5
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Laplace transform operator, s-8, e
and f (aq) is a function of the indicated difference between top and bottom halves of detectors of the power range neutron ion chambers; with gains to be selected based on measured instrument response during plant STARIUP tests such that:
(i) for q q
between -35% and +10K, f (aq) = 0, where q, and qb P'"t g
b RATED THERMAL POWER in the top and bottom halves of the core respectively, and qt+qb is total THERMAL POER in percent of RATED THERMAL POWR, y
(ii) for each percent that the magnitude of q q exceeds -35%, the M-16 Trip t
b Setpoint shall be automatically reduced by 1.2'% of its value at RATED THElut4L POW R, and J. 22.
I.o# 5f o2.
tlii) for each percent that the magnitude of q
~9 t
b exceeds +101, the N-16 Trip Setpoint shall be automatically reduced by 4-465-of its value at RATED THERMAL
' *0MER-I* %
r. s s ss o's NoiE 2:
The channel's maximum Trip 5etpoint'shall not exceed its computed Trip setpoint by more than 1.4%
of span.
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III-WM-St'f 80 AffACM(Ni 2 PAGE 19 0F 20 LIMITING SAFETY SYSTEM SETTINGS f
BASES Undervoltace and Underfrecuency - Reactor Coolant Pump Busses The Undervoltage and Underfrequency Reactor Coolart Pump Bus trips pro-vide core protection against DNB as a result of complete loss of forced coolant.
flow.
The specified setpoints assure a Reactor trip signal is generated before the Low Flow Trip Setpoint is reached.
Time delays are incorporated in the Underfrequency and Undervoltage trips to prevent spurious Reactor trips from momentary electrical power transients.
For undervoltage, the delay is set so that the time required for a signal to reach the Reactor trip breakers following the simultaneous trip of two or more reactor coolant pump bus circuit breakers shall not exceed 1.2 seconds.
For underfrequency, the delay is set so that the time required for a signal to reach the Reactor trip breakers after the Underfrequency Trip Setpoint is reached shall not exceed 0.3 second.
On decreasing power the Undervoltage and Underfrequency Reactor Coolant Pump Bus trips are automatically blocked by P-7 (a power level of approximately 10%
of RATED THERMAL POWER with a turoine first stage chamber pressure at approximately 10% of full power equivalent); and on increasing power, reinstated automatically by P-7.
Turbine Trio
{
A Tu'rbine trip initiates a Reactor trip.
On cecreasing power the Reactor trip from the Turbine trip is autom?tically blocked by P-7 (a power level of approximately 10% of RATED THERMAL POWER); and on increasing power, reinstated automatically by P-7.
Safety injection Inout frem ESF If a Reactor trip has not already been generated by the Reactor Trip System instrumentation, the ESF automatic actuation logic channels will initiate a Reactor trip upco any signal which initiates a Safety Injection.
The ESF instrumentation channels which initiate a Safety Injection signal are shown in Table 3.3-1, e d.
1 CCMANCHE PEAK - UNIT 1 8 2-7
Tu-tMeo AmamM 3 CPSES Technical Specifications N M c5 IL NRC Oraft 2 Markup Section 3/4.0 Change 10#
Justification for Change 5104 As discussed with the NRC October 5, 1988, the surveillance requirements of the steam generator Technical Specification, previously planned for relocation to the CPSES Technical Specification Improvement Program, are now proposed for relocation to Specification 4.0.6.
Removal of the LCO is based on a risk survey performed by TV Electric.
5105 Modified Table 4.0-1 notation to more clearly explain that in the second and third inspections, only one steam generator has to be inspected.
The way it is presently written implies that both the steam generators not inspected in the first inspection nave to be inspected in the second inspection and again in the third inspection.
5106 This cht9ge makes the Bases read the same as the Specification.
There are numerous examples of surveillance requirements that cannot be performed during the Operational Modes or other conditions for which the requirement of the LC0 apply.
A few examples would be Blackout Testing, Safety Injection Actuation Testing', etc.
Trr4MH B31Bo ATTACHMOT3 PAGE 6 0F 12 APPLICABil.!TY SURVEILLANCE REQUIREMENTS (Continued) b.
Surveillance intervals specified in Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda for the inservice inspei. tion and testing activities required by the ASME Boller and Pressure Vessel Code and applicable Addenda shall be applicable as follows in these Tira nical Spscifications:
ASME Soiler and Pressee Vessel Required frequencies for Code and applicable Addenda performing inservice terminology for inservice inspection and testing inspection and testino activities activities Weekly At least once per 7 days Monthly At least once per 31 days Quarterly or every 3 months At least once per 92 days Samiannually or every 6 months At least once per 184 days Every 9 months At least once per 276 days Yearly or annually At least once per 366 days c.
The provisions of Specification 4.0.2 are applicable to the above required frequencies ft nerforming inservice inspection and testing activities; J
d.
Performance of the above inservice inspection and testing activities l
shall be in addition to other specified Surveillance Requirements; and e.
Nothing in the ASME Boiler and Pressure Vessel Code shall be construed l
to supersede the requirements of any Technical Specification, g
k. 6,(c k N e illa.* c c Re9airem'ds fo r-e_%h ste m pivder d a.O b e.
puSermed felleW.A) tkt a qmukd lut.rvic s A sp ech'm p roye spec' fled be.\\ow ud Akt repletmwh or $pec; f u NE Y.C.f.
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CCMANCHE PEAK - UNIT 1 3/4 0-3
T u Btito Sntec4mW 3 Phra 6 A er it fd 5 E e.T Fon PAos 3/4 o-3 1.s.s vos q,g,9 -4r441 Steam Generator Sample Selection and Inspection - Each steam generator l
shall be determined OPERA 8LE during shutdown by selecting and inspecting at
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1 east the minimum number of steam generators specified in Table 4./'-1.
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d.C.(c
t. t. 5. 2 Steam Generator Tube Sample Selection and Inspection - The steam l
generator tube minimum sample size, inspection result classification, and the corresponding action required shall be as specified in Table 4.f-2.
The j o l
inservice inspection of steam generator tubes shall be performed ~at tne fre-quencies specified in Specification 4r443 and the inspected tubes shall be ' 4.0.k l verified acceptable per the acceptance criteria of Specification-4r4r W.
The l
tubes selected for each inservice inspection shall include at least N of all the expanded tubes and at least 3% of the remaining number of tubes in all steam generators; the tubes selected for these inspections shall be selected on a random basis except:
a.
Where experience in similar plants with similar water chemistry indicates critical areas to be inspected, then at least 50% of the tubes inspected shall be from these critical areas; b.
The first sample of tubes selected for each inservice inspection (subsequent to the preservice inspection) of each steam generator shall include:
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(, B o r it s, g. A r. i o 4 TYR NAMTSYSTEM e, TEAM OCXCRATMS-t SURVEILLANCE REQUIREMENTS (Continued) 1)
All nonplugged tubes that previously had detectable wall penetrations (greater than 20%),
2)
Tubes in those areas where experience has indicated potential problems, and
4. 0. G 3)
A tube inspection (pursuant to Specification +4 '.r.44.8) shall be performed on each selected tube.
If any selected tube does not permit the passage of the eddy current probe for a tube inspection, this shall be recorded and an adjacent tube shall be selected and subjected to a tube inspection.
c.
The tubes selected as the second and third samples (if required by Table 4.1-2) during each inservice inspection may be subjected to a partial tube inspection provided:
1)
The tubes selected for these samples include the tubes from those areas of the tube sheet array where tubes with imperfections were previously found, and 2)
The inspections include those portions of the tubes where I
imperfections were previously found.
The results of each sample inspection shall be classified into one of the following three categories:
Ca, teaory Inspectirn Results C-1 Less than 5% of the total tubes inspected'are degraded tubes and none of the inspected tubes are defective.
C-2 One or more tubes, but not more than 1.% of the total tubes inspected are defective, or between 5% and 10% of the total tubes inspected are degraded tubes.
C-3 More than 10% of sne total tubes inspected are degraded tubes or more than 1% of the inspected tubes are defective.
Note:
In all inspections, previously degraded tubes must exhibit significant (greater than 10%) further wall penetrations to be included in the above percentage calculations.
COMANCHE PEAK - UNIT 1 3/4-4,
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-4EAC40" COOLP' SYSTEM--
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--GTEAM-GENEAATG45--
SURVE!LLANCE REQUIREMENTS (Continued) y,c,y ' '.5r3 Inspection Frequencies - The abovv required inservice inspections of steam generator tubes shall be performed at the following frequencies:
The first inservice inspection shall be performed af ter 6 Effective a.
Full Power Months (EFPM) and before 12 EFPM and shall include a special inspection of all expanded tubes in all steam gen 6rators.
Subsequent inservice inspections shall be performed at intervals of not less than 12 nor more than 24 calendar months after the previous inspection.
If two consecutive inspections, not including the preser-vice inspection, result in all inspection results falling into the C-1 category or if two consecutive inspections demonstrate that previously observed degradation has not continued and no additional degradation has occurred, the inspection interval may be extended to a maximum of once per 40 months; b.
If the results of the inservice inspec ion of a steam generator conducted in accordance with T6bic 4.V2 at 40-month intervals fall in Category C-3, the inspection frequency shall be increased to at least once per 20 months.
The increase in inspection frequency shall apply until the subsequent inspections satisfy the criteria of Specification 4r4,4.3a.; the interval may then be extended to a maximum of cnceper 40 months; and
'{. 0. h c.
Additional, unscheduled inservice inspections shall be performed on each steam generator in accordance with the first sample inspection specifiedinTable4.\\Qduringtheshutdownsubsequenttoanyof the following conditions:
o 1)
Primary-to secondar" tubes leak (not including leaks originating from tube-to tube sheet welds) in excess of the limits of i
l Specification 3.4.6.2, or 2)
A seismic occurrence greater than the Operating Basis Earthquake, or 3)
A loss of-coolant accident requiring actuation of the Engineered Safety Features, or l
4)
A main steam line er feedwater line break.,
l l
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COMANCHE PEAK - UNIT 1 3/4 tir 05 L
T11-BS160 hTTBLR m( OT 3 f 44 Sici 7)AGG Gb of IL APpuCAGILifr RCACTOR COOLANT SYSTEM--
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-0TCA" OEMERATOR SURVEILLANCE REQUIREMENTS (Continued)
I 4.0. 0 4 e.L4 Acceptance Criteria a.
As used in this specification:
1)
Imperfection means an exception to the dimensions, finish, or I
contour of a tube from that required by fabrication drawings or i
specifications.
Eddy-current testing indications Lelow 20% of the nominal tube wall thickness, if detectable, may be considered as imperfections; 2)
Dearadation means a service-induced cracking,
wastage, wear, or general corrosion occurr:ng on either inside or outside of a tube; 3)
Degraded Tube means a tube containing imperfections greater tnan or equal to 20% of the nominal wall thickness caused by degradation; I
4)
% Oearadation means the percentage of the tube wall thickness affected or removed by degradation; I
5)
Defect means an imperfection of such severity that it exceeds the plugging limit. A tube containing a defect is defective; 6)
Pluaaina limit means the imperfection depth at or beyond which the tuot shall be removed from service and is equal to 40%*
of the nominal tube wall thickness; 7)
Unserviceable Jescribes the condition of a tube if it leaks or contains a defect large enough to affect its structural integ-rity in the event of an Operating Basis Earthquake, a loss-of-coolant accident, or a steam line or feedwater line break as specified in Specification-4r44.3c., above; 4.00 8)
Tube Inspection means an inspection of the steam generator tube from the point of entry (hot leg side) completely around the U-bend to the top support of the cold leg; and i
"Value to be determined in accordance with the recommendatiors of Regulatory Guide 1.121 August 1976.
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COMANDiE PEAK - UNIT 1 3 /4 -+-FE 0-6
Txx tt W e nwne.m r 3
r. 4 0 's to4 Mu 6 s. u t?.
h?NCA6 m T*Y
-4EAC4M-4006 ANT-Gy& TEM-iui
-GTEAM4ENEAATM-
$URVE!Lt.ANCE REQUIREMENTS (Continued) 9)
Preservice Inspection means an inspection of the full length of each tube in eacn steam generator performed by eddy current techniques prior to service to establish a baseline condition of the tubing.
This inspection shall be performed prior to initial POWER OPERATION using the equipment and techniques expected to be used durir.g subsequent inservice inspections.
b.
The steam generator shall be determined OPERA 8LE af ter completing the corresponding actions (plug all tubes exceeding the plugging limit and all tubes containing through-wall cracks) required by Table 4./-2.
O z/, C,4,
-4r4rb5 Reports a.
Within 15 days following the completion of each inservice inspection of steam generator tubes, the number of tubes plugged in each steam
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generator shall be reported to the Commission in a Special Report pursuant to Specification 6.9.2;
~
b.
The complete results of the steam generator tube inservice inspec-tion shall be submitted to the Commission in a Special Report pur-I suant to Specification 6.9.2 within 12 months following the completion of the inspection.
This Special Report shall include:
1)
Number and extent of tubes inspected, 2)
Location and percent of wall thickness penetration for each indication of an imperfection, and 3)
Identification of tubes plugged.
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Results of steam generate-tube inspectiont which fall into Category z
C-3 shall be reported to the Cossaission pursuant to 10 CFR feet 50.72$1) within four hours of init'il discovery, andVpursuant to Specifica-g 7 tion 6?/.2 within 30 days and prior to resumption of plant operation.
d,I.his f oort shall provide a description of investigations conducted h
to determine cause of the tube degradation and corrective measures Sp taken to prevent recurrence.
COMANCHE PEAK - UNIT 1 3 /4 4-N-0-7
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TABLE 4./-1
$4 2
de g
MINIMUM !nBSER OF STEAM GENERATORS TO BE x
.i" x
INSPECTED DURING INSERVICE INSPECTION m
w V
oT n
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Preservice Inspection Four zZ Mo. of Steam Generators per Unit four w
First Inservice Inspection Two Second & Subsequent Inservice Inspections One8 (cu_ 6 euk in, ped.A pte;o l*
TA8tE NOTATIONS c
T b4 5tos s.e ik.t w e r c.
Exi ;f b etenee-two steam generators 6not inspected during !the first inservice inspection /
D 1.
o9 shall be inspected during the second and third inspections / For the fourth and subsequent d sa inspections, the inservice inspection may be limited to one steam generator on a rotating T
schedule encompassing 12% of the tubes if the results of the -f 6 t = previous inspections ofr Lk gu,. du indicate that all steam generators are performing in a like manner. Note that under some
,,g,3 circumstances, the operating conditions in one or more steam generators may be found to be more severe than those in other steam generators. Under such circumstances the sample sequence shall be modified to inspect the most severe conditions.
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TABLE 4./-2
E $
h* 3 e STEAM GENERATOR TURE INSPfCTION o
-t pu O
l',T SAMPLE INSPECil0N 2FA SAMPLE INSPECTION 3RD SAMPLE INSPECTION z
Q$ ample Size Re' alt Action Required Result Action Required Result Action Required
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yA minimum of C-1 None N.A.
N.A.
N.A.
N.A.
g5 Tubes per
,5.G.
C-2 Plug defective tubes C-1 None N.A.
N.A.
and inspect additional l
c5 25 t h s in this 5.G.
Plug defective tubes C-1 None C-2 and inspect additional 45 tubes in this 5.G.
C-2 Plug defective ti&es Perform action for C-3 C-3 result of first sample Perform action for C-3 C-3 result of first N.A.
N.A.
,g sample
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C-3 Inspect all ti&es in All other l
l 0o this 5.G., plug de-5.G.s are None N.A.
N.A.
l fective tubes and C-1 inspect 25 tubes in each other 5.G.
Some 5.G.s Perform action for N.A.
M.A.
I C-2 but no C-2 result of second I
additional sample 1
Notification to NRC 5.G. are pursuant to MG C-3
-(b)(2)-of-14-GA-Pert 50 Additional Inspect all tubes in
/0 CFC CD.*7A(bg) 5.G. is each 5.G. and plug g
C-3 defective tubes.
l Notification to NRC N.A.
N.A.
pursuant to ".,0.72 ib)(2) Of le CFA-
-Pert SS /OCf( 72(b)(A) 1 5=
idhere n is the number of steam generators inspected during an inspection
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III-( % t te l9.o AffACMDT3 PAGE 10 0F 12 APPLICASILITY b
BASES "Surveillance requirements are requirements relating to test, calibration, or inspection to ensure that the necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the limiting conditions of operation will be met."
7.g3 S ic t.
Specification 4.0.1 establishes the requirement that surveillances must be nie f wrfere:c during the OPERAT]CNAL H0 DES or other conditions for =nich the re-quirements of the Limiting Conditions for Operation apply unless otherwise stated in an individual Surveillance Requirement.
The purpose of this specifi-cation is to ensure that surveillances are performed to verify the operational status of systems and components and that parameters are within specified limits to ensure safe operation of the facility when the plant is in a MODE or other specified condition for which the associated Limiting Conditions for Operation are applicable.
Surveillance Requirements do not have to be performed when the facility is in an OPERATIONAL MODE for which the reavirements of the asse-ciated Limiting Condition for Operation do not apply unless otherwise specified.
The Surveillance Requirements associated with a Special Test Exception are only applicable when the Special Test Exception is used as an allowable excep-tion to the requirements of a specification.
Specification 4.0.2 establishes the conditions under which the specified time interval for Surveillance Requireaents may be extenced.
Item a. permits an allowable extension of the normal surveillance interval to facilitate surveil
lance scheduling and const e ration of plant operating conditions that may not be suitable for conducting tne surveillance; e.g., transient conditions or other ongoing surveillance or maintenance activities.
Item b. limits the use of the provisions of item a. to ensure that it is not used repeatedly to extend the surveillance interval beyond that specified.
The limits of Specification 4.0.2 are based on engineering judgment and the recognition that the most prob-ab9 result of any particular surveillance being performed is the verificaion of conformance with the Surveillance Requirements.
These provisions are suf-ficient to ensure that the reliability ensured through surveillance activities is not significantly degraded beyond that obtained from the specified survell-lance interval.
Specification 4.0.3 establishes the failure to perform a Survei11ance Require-ment within the allowed surveillance interval, defined by the provisions of Specification 4.0.2, as a condition that constitutes a failure to meet the OPERABILITY requirements for a Limiting Condition 'cr Operation.
Under the provisions of this specification, systems and c.,
sents are assumed to be OPERABLE when Surveillance Requirements have bee'.atisfactorily performed within the specified time interval.
However, nothing in this provision is to be construed as implying that systems or components are OPERABLE when they are fout c or known to be inoperable although still msting the Surveillance Require-ments.
This specification also clarifies that the ACTION requirements are applicable when Surveillance Requirements have not been completed within tne I
allowed surveillance interval and that the time limits of the ACTION require-ments apply from the point in time it is ice,tified snat a surveillance has not been performed and not at the time that the allowed surveillance interval CCMANCHE PEAK - UNIT 1 B 3/4 0-4
Int 66Mt$4780 i
ATTACWOf3 PAGE12of12 APPLICA8!LITY BASES A t dre When a e%4-is required to comply with ACTION requirements, the provisions of g,
Specification 4.0.4 do not apply because this would delay placing the facility in a lower MODE of operation.
Sp eification 4.0.5 establishet the requirement that inservice inspection of A3ME Coce class 1, 2, and 3 cosponents and inservice testing of ASME Coce Class 1, 2, and 3 puros and valves shall be performed in accordance with a periodically updated version of Section XI of the ASME Soiler and Pressure Vessel Code and Addenda as required by 10 CFR 50.551.
These requirements apply except when relief has been provided in writing by the Commission.
This specification includes a clarification of the frequencies for performing the inservice inspection and testing activities required by Section XI for the ASME Boiler and Pressure vessel Code and applicable Addenda. This clari-fication is provided to ensure consistency in surveillance intervals througn-out the Technical Specifications and to remove any ambiguities relative to the frequencies for performing the required inservice inspection and testing activities.
Under the terms of this specification, the more restrictive requirements of the Technical Specifications take precedence over the ASME toiler and Pressure Vessel Code and applicable Addenda. The requirements of Specification 4.0.4 to perform surveillance activities tefore entry into an CPERATIONAL MODE or I
other specified condition takes precedence over the ASME toiler and Pressure vessel Code provision which allows pumps and valves to be tested up to one week after return to normal operation.
The Technical Specification definition of OPERA 8LE does not allow a grace period before a component, that is not capable of performing its specified function, is declared inoperable and takes precedence over the A$ME toller and Pressure Vessel Code provision which allows a valve to be incapable of performing its specified function for un to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months before being declared inoperable.
T. t>4 510 4 Spec. k m v.o.<, (r e r)
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CCMANCHE PEAK - UNIT 1 8 3/4 0-6
rn-tM M MTmc.omer 3 tem st A o f 13 I.. b. O s t e 4 TMstit1" P'oc. PA M 63/4 0-(s fr.sMMik o The Surveillance Requirements for inspection of the steam generator tubes ensure that the structural integrity of this portion of the RCS will be main-tained. The program for inservice inspection of steam generator tubes is based on a modification of Regulatory Guide 1.83, Revision 1.
Inservice inspection of steam generator tubing is essential in order to maintain surveillance of the conditions of the tubes in the event that there is evidence of mechanical damage or progressive degradation due to design, manufacturing errors, or inservice conditions that lead to corrosion.
Inservice inspection of steam generator tubing also provides a means of characterizing the nature and cause of any tube degradation so that corrective measures can be taken.
Selected tubes in the preheater section of each 04 and 05 steam generator have been modified to correct the tube vibration degradation phenomenon esperi-enced by certain Westinghouse steam generators.
The modification consisted of
)
expanding these tubes in the vicinity of the support plates and is designed to Ilmit the amplitude of vibration.
These expanded tubes are subject to a special inspection whenever the steam generators are opened for inservice eddy current testing.
)
The plant is expected to be operated in a manner such that the secondary coolant will be maintained within those chemistry limits found to result in negligible corrosion of the steam generator tubes.
If the secondary coolant i
chemistry is not maintained within these limits, localized corrosion may likely i
result in stress corrosion cracking.
The extent of cracking during plant operation would be limited by the limitation of steam generator tuce leakage l
between the Reactor Coolant System and the Secondary Coolant System (reactor-to secondary leakage = 500 gallons per day per steam genarator and a total leakage of 1 GPM to all steam generators).
Cracks having a reactor-to-i t
secondary leakage less than this limit during operation will have an adequate margin of safety to withstand the loads imposed during normal operation and by postulated accidents.
Operating plants have demonstrated that reactor-to-secondary leakage of 500 gallons per day per steam generator can reaoily bc l
detected by radiation monitors of steam generator blowdown.
Leakage in excess of this limit will require plant shutdown and an unscheduled inspection, l
during which the leaking tubes will be located and plugged.
Wastage type defects are unlikely with proper chemistry treatment of the i
secondary coolant. However, even if a defect should develop in service, it will be found during scheduled inservice steam generator tube examinations.
Plugging will be reoutred for all tubes with imperfections exceecing the i
plugging limit of 40% of the tube nominal wall thickness.
Steam generator tube inspections of operating plants have demonstrated the capability to reliably detect degradation that has penetrated 20% of the original tube wall l
thickness,g g
gg
)
Whenever the results of~ any steam generator tubing intervice inspection fall into Category C-3, these results will be reported to the Coanission l
pursuant to 10 CFR 50.72 within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months from initial discovery an49urst. ant to Specification 6.9.2 within 30 days and prior to resumption of l
plant operation.
Such cases will be considered by the Commission on a case-by-case basis and may result in a requirement for analysis, laboratory examina-tions, tests, additional eddy-current inspection, and revision of the Technical Specifications, if necessary.
l 3-Ttv M76c>
Pn%camuT 4 CPSES Technical Sp2cifications
?Mt M oF 29 NRC Oraft 2 Markup Section 3/4.1 Change 10#
Justification for Change 5107 Adds the exclusion of Specification 4.0.4 since the actual overall core reactivity balance required to perform the comparisen cannot be performed until after criticality is i
obtained.
This situation will occur after every refueling i
since the changing out of fuel will invalidate the 1,
previous surveillance test.
5108 Change the surveillance requirement wording back to the standard technical specifications.
This is based on an i
operational constraint that was overlooked when the change l
was initially submitted.
The operational constraint comes about when the plant is in a solid condition in Mode 5.
i It is not very desirable, from a safety concern of having the potential of overpressurizing the RCS, to require charging at 120 gpm into a solid plant.
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t TittMtt96180 AliAC2(WI 4 PAtt 5 0F 29 REACTIVITY CONTROL SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)
When in MODE 3 or 4 at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by consideration of e.
the following factors:
1)
Reactor Coolant 5ystem boron concentration, 2)
Control red position.
3)
Reactor Coolant System average temparature, 4)
Fuel burnup based on gross thermal energy generation, 5)
Xenon concentration, and 6)
Samarium concentration.
4 4.1.7.1.2 The overall core reactivity bilance shall be compared to predi:ted values to demonstrate agreement within 21% ak/k at least once per 3. Effective Full Power Days (EFPD).
This comparison shall consider at least those factors stated in Specification 4.1.1.1.le., above.
The predicted reactivity values shall be aojusted (normalized) to correspond to l'
the actual core conditions prior to exceeding a fuel burnup of 60 EFP0
(
after each fuel loading. TA gre m r e 5. Sp u : 5 :c a tin 4. C. 9/
m 4 r e.
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be. E 3\\ EFPL fell.aia) initLI f w I l o u l o r r e -f. a.l y,
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l CCMAl!CHE PEAK - UNIT 1 3/4 1-2
T11 SeHf 8B'183 Ai!ACR4 NI 4 iAG(12Cf29 I
REACT!V!TY CONTROL SYSTEMS l
CHARGING PUMP - $HUTDOWN LIMITING CON 0! TION FOR OPERATION 3.1.2.3 At least one charging pump in the boron injection flow path required by Specification 3.1.2.1 shall be OPERA 8LE and capable of being powered from an OPERA 8L2 emergency power source.
APPLICABILITY:
MODES 5 and 6.
ACTION:
With no charging pump OPERABLE or capable of being powered from an OPERABLE emergency power source, suspend all operations involving CORE ALTERATIONS or positive reactivity changes.
1
$URVE!LLANCE REQUIREMENTS pes',h:n. d'.spW.a m sak 4.1.2.3.1 At least once per 92 days the above required charging pump shall be 4
demonstrated OPERABLE by verifying that the flow path required by Specifica-tion 3.1.2.la is capable of delivering at least 30 gem to the RC5; or z.b.4 Sie?
4.1.2.3.2 - At 1;;;t :n:; p;r 92 d:y: by v:rifying-t u t-tM f 6 p;th r;; tr;d-by 0;;;ificetier 3.1.2.1t i; ;;;;;;ie of delivering et 1;;;t-120 ge;-te-the RCS.
4.1.2.3.3 A maximum of two charging pumps shall be OPERABLE, one charging pump shall be demonstrated inoperable
at least once per 31 days, except when the reactor vessel head is removed, by verifying that the motor circuit breakers are secured in the open position.
ctatr'dqd The aoove required charging pump shall be demonstrated OPERA 8LE by i
verifying, on recirculation flow that a differential pressure across the pump of greater than or equal to[ arnl,psid is developed when tested pursuant to Specification 4.0.5.
"An inoperable pump may be entrgized for testing provided the discharge of the pump has been isolated from the RCS by a closed isolation valveMwith power removed from the valve operator $br by a manual isolation valve $ secured in the closed position.
gg CCMANCHE PEAK - UNIT 1 3/4 1-9
III-M M 8 876 0 AffACM(NT 4 PAE 13 0F 29
{k l
REACTIVITY CONTROL SYSTEMS I
CHARGING PUMPS - OPERATING LIMITING CON 0! TION FOR OPERATION ccvde 6 ha.3J 3.1.2.4 At least two charging pumps shall be OPERABLE.
3 APPLICA8ILITY: MODES 1, 2, 3, and 4.#
ACTION:
With only one charging pump OPERA 8LE, restore at least two charging pumps to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT STAN08Y and berated to a SHUT 00WN MARGIN equivalent to at least 1% ak/k at 200*F within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ; restore at least two charging pumps to OPERABLE status within the next 7 days or be in COLD SHUT 00WN within the next 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .
SURVEILLANCE REQUIREMENTS 4.1.2.4.1 The required centrifugal charging pump (s) shall be demonstrated OPERA 8LE by testing pursuant to Specification '.1.2.3.2. 4c4 IU 4.1.2.4.2 The required positive displacement charging pump shall be demonstrated OPERABLE by testing pursuant to Specification '.1.2. 3.1. 'l.l.2..ic.. l g.
4.1.2.4.3 WheneverthetemceratureofoneormoreoftheReacto[ Coolant System (RCS) cold legs is less than or equal to 350'F, a maximum of two charging pumps shall be OPERABLE, one chargir.g pump shall be demonstrated inoperable
at least once per 31 days by verifying that the motor circuit breakers are secured in the open position.
4 An inoperable pump may be energi:ed for testing provided the discharge of g
the pump has been (selated from the RCS by a closed isolation valveekith B
power removed from the valve operato6')or by a manual isolation valvepsecured 3 in the closed position.
?
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1:1Ct6 S
CCMANCHE PEAK - UNIT 1 3/4 1-10
III1)Mit 84186 AllACHMENT 4 PAGE 14 0F 19 REACTIVITY CONTROL SYSTEMS 80 RATED WATER SOURCE - SHUTDOWN LIMITING CON 0! TION FOR OPERATION 3.1.2.5 As a minimum, one of the folio.ving borated water sources shall be OPERA 81.E:
a.
A boric acid storage tank with:
1)
A minimum contained borated water volume of (6385) gallons,
([Later]% of span), when using the boric acid transfer pump.
2)
A minimum contained borated water volume of 15,123] gallons I
o.
((Later]% of span), when using the gravity feed connection.
3)
A minimum boron concentration of 7000 ppm and 4)
A minimum solution temperature of 65'F.
b.
The refueling water storsge tank (RWST) with:
1)
Aminimumcontainedboratedwatervolumeof[101,120] gallons, g'
([Later]% of span),
i 2)
A minimum b'oron concentration of 2000 ppe and 3)
A minimum solution temperature of 40*F.
APPLICABILITY: MODES 5 and 6.
ACTION:
With no borated water source OPERABLE, suspend all operations involving CORE ALTERATIONS or positive reactivity changes.
SURVEILLANCE REQUIREMENTS 4.1.2.5 The above required borated water source shall be demonstrated OPERA 8LE:
a.
At least once per 7 days by:
1)
Verifying the boron concentration of the water, 2)
Verifying the contained borated water volume, and 3)
Verifying the boric acid storage tank solution temperature when it is tht source of borated water.
b.
At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by verifying the RWST temperature when it l/
is the source of borated water and tne outside air temperature is less than 40'F.
CCMANCHE PEAK - UNIT 1 3/4 1-11
TH 9664 66160 AITAcmENT 4 fkSE270F2%
DP..A.U.
REACTIVITY CONTROL SYSTEMS s
BASES N00ERATOR TEMPERATURE COEFFICIENT (Continued) involved subtracting the ir.cremental change in the MOC associated with a core condition of all rods inserted (most positive MOC) to an all rods withdrawn condition and, a conversion for the rate of change of moderator density with temperature at RATED THERMAL POWER conditions.
This value of the MOC was then transformed into the limiting HT value -4.0 x 10 4 ok/k/'F.
The MTC value of -3.1 x 10 4 ak/k/*F represents a conservative value (with corrections for burnup and solubic boron) at a core condition of 300 ppm equilibrium boron conce:tration and is obtained by making these corrections to the limiting MTC value of -4.0 x 10
ak/k/'F.
The Surveillance Requirements for measurement of the MTC at the beginning and near the end of the fuel cycle are adequate to confirm that the MTC remains within its limits sirce this coefficient changes slowly due principally to the 4
reduction in RCS boron concentration associated with fuel burnup.
3/4.1.1.4 MINIMUM TEMPERATURE FOR CRITICALITY g
This specification ensures that the reactor will not be made critical with the Reactor Coolant System average temperatt e less than 551*F. This limitation is required to ensure:
(1) the moderator temperature soefficient is within it analyzed temperature range, (2) the trip instrumentation is within its normal operating range, (3) the pressuaizer is capable of being in an CPERABLE status with a steam bubble, and (4) the reactor vessel is above its minimum RT temperature.
NOT 3/4.1.2 BORAT!0N SYSTEMS The Boron Injection System ensures tnat negative rea:tivity control is available during each mode of facility operation.
The components recuired to perform this function include:
(1) borated water sources, (2) charging puirps, (3) separate flow paths. (4) boric acid transfer pumps, and (5) an emergency power supply from OPERA 8LE diesel generators.
a With the RCS average temperature above 200'F, a minimum of two boron injection flow paths are required to ensure single functional capanility in the event an assumed failu're renders one of the flow paths inoperable. The boration capability of either flow path is sufficient to provice a SHUTDChN MARGIN from expected operating conditions of 1.6% Ak/k after xenon decay and cooldown to 200*F.
The maximum expected boration capability requirement occurs at E0L from full power equilibrium xenon conditions and requires
/4.ccc (22,070] gallons of 7000 ppm borated water from the boric acid storage tanks or g'oco-{49r966 gallons of 2000 ppm borated water from the refueling water storage
(
tank (RWST).
4 CCMANCHE PEAK - UNIT 1 B 3/4 1-2
III 9$He 44700 Af f ACMDi 4 '
PAE 26 0F 29 REACT!V!TY CONTROL SYSTEMS w ' '
BASES 80 RATION SYSTEMS (Continued)
With the RCS temperature below 200*F, one loren Injection System is acceptable without single failure consideration on the basis of the stable reactivity condition of the reactor and the additional 'estrictions prohibiting CORE ALTEPATIONS and positive reactivity changes in the event the single loron Injection System becomes inoperable.
The limitation for a maximum of two charging pamps to be OPERA 8LE and the d
requirement to verify one charging pump to be inoperable below 350*F provides j
assurance that a mass addition pressure transient can be relieved by the operation of a single PORV.
I The limitation for minimum solution temperature of the borated water sources are sufficient to prevent boric acid crystallization with the highest allowable boron concentration.
i l
The boron capability required below 200*F is sufficiant to provide a i
SHUT 00WN MARGIN of 15 Ak/k after xenon decay and cooldown from 200*F to 140*F.
This condition requires either [0,101] gallons of 7000 ppe borated waterfromtheboricacidstoragetankpsr (101,100) gallons of 2000 ppe borated water from the RWST.
/,/co G,voo The contain'ed water volume limits include allowance for water not available because of discharge line location and other physical charac* eristics.
l The Itaits on contained water voluee and baron concentration of the RWST j
are also consistent with Specification 3.5.4.
The OPERA 81LITY of one Boron Injection System during REFUELING ensures that this system is available for reactivity control while in MODE 6.
i
)
3/4.1.3 MOVA8LE CONTROL A55EMBL!f5 i
(1) acceptable power distri-The specificati ss of this section ensure that:
bution limits are maintained, (2) the minimum SHUTOOWN MARGIN is maintaineo, and (3) the potential effects of rod misalignment on associated accident analyses are limited.
OPERASILITY of the control rod position indicators is rc 'Jired to l
determine contrcl rod positions and thereby ensure compliance with the control rod alignment and insertion limits. Verification that the Digital Rod Position Indicator agrees with the demanded position within 12 steps at 24, 48, 120, and 228 steps withdrawn for the Control Banks and 18, 210, and 228 steps with-drawn for the Shutdown tanks provides assurances that the Digital Rod Position Indicator is operating correctly over the full range of indication.
Since the Digital Rod Position Indication System does not indicate the actual shutcown red l ',
position between 18 steps and 210 steps, only points in the indicateo ranges 1
are picked for verification of agreement with demanded position.
COMANCHE PEAK - UNIT 1 B 3/4 1-3 i
._.__,-,_~_.___....___..r_,,
~17(x. 6878o ATincamcc 5 CPSES Technical Specifications Pact 4 A er 15 NRC Draft 2 Markup i
Section 3/4.2 Chanae 108 Justification for Chance 5109 The note that allows 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months of operation outside the AFD target band needs to be tied to the LCO, not to part of an Actinn Statement. With the way that the note is presently referenced, the surveillance testing of the power range neutrcn flux channels being tested pursuant to t
Specification 4.3.1.1 would not be able to be tested since it is guaranteed that when indu'.Ing the xenon transient, AFD will be outside the target band for an extended amount of time.
By followirg the Action Statement, it would be required to reduce power below 50% power which would prevent the completion of the surveillance requirement 4.3.1.1 which requires power to be above 50%.
i l
2
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1
+
1
til MHt erttp AfIACV N 5 PAGE 5 # 25 3/4.2 POWER 0!$TRIBUT!0N LIMITS 3/4.2.1 AXIAL FLUX O!FFERENCE LIMITING CONDJTION FOR OPERATION 3.2.1 The indicated AXIAL FLUX DIFFERENCE (AFD) shall te maintained within the following target band (flux difference units) about the target flux difference:
15% for core average accumulated burnup of less than or equal to a.
3000 WO/MTU; and b.
+ 3%
12% for core average accumulated burnup of greater than 3000 WO/MTU.
7. p t Stoi The indicated AFD may deviate outside the above required target band at greater than er aqual to 50% but less than 90% of RATED THERMAL POWER provided the inci-cated AFD is within the Acceptable Operation Limits of Figure 3.2-1 and the cumu-lative penalty deviation time does not exceed I hou/ huring the previous 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .
The indicated AFD may deviate catside the above required target band at greater than 15% but less than 50% of RATED THERMAL POWER provided the cumulative penalty deviation time does not exceed I hour during the previous 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .
APPLICABILITY:
MODE 1, above 15% of RATED THERMAL POWER.*
ACTION:
With the indicated AFD outside of the above required target band and a.
(
with THERMAL POWER greater than or equal to 90% of RATED THERMAL POWER, within 15 minutes either:
I 1.
Restore the indicated AFD to within the target band limits, or 2.
Reduce THERMAL POWER to less than 90% of RATED THERMAL POWER.
b.
With the indicated AFD outside of the above required target band for more than 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months of cumulative penalty deviation time during the previous 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or outside the Acceptable Operation Limits of Figure 3.2 1 and with THERMAL POWER less than 90% but equa) to or greater than 50% of flATED THERMAL POWER:
1.
Reduce THERMAL POWER to less than 50% of RATED THERMAL POWER within 30 sinwis, and 2.
Reduce the Power Range Neutron Flux - High Trib5etpoints to less than or equal to 55% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .
,,"See Special Test Exceptions Specification 3.10.2.
Surveillance testing of the Power Range Neutron Flux Channels may be perforsed pursuant to Specification 4.3.1.1 proviced the indicated AFD is maintained within l
the Acceptable Operation Limits of Figure 3.2 1.
A total of 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months operatinn may be accumulated with the AFD outside of the above required target band during testing without penalty deviation.
COMANCHE PEAK - UNIT 1 3/4 2-1 l
L
1n MH2 BA T EC AllACHrDI5 y?\\
PCsER O!$TRiBUTION, LIMITS LIMITING CON 0! TION FOR OPERATION ACTION (Continued) b.
Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months of initially being outside the above limits, verify through intore flux mapping a*4-4GS-Aet:1-f1;w rate-ee*peehen that b= " '"
in; c Wa444+*-ef-4-4*d-RGi-toubMew-eate-are-re stored to wi thin the above limits, or reduce THERMAL POWER to less than 5% of RAT!0 a
THERMAL POWER withi* the next 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months p d 3
Identify and correct the cause of the out-ofelimit condition prior c.
to increasing THERMAL POWER above the reduced THERMAL P0hER limit r* Quired by ACTION a,2. And/or b.,
above; subsequent POWER OPERATICN may proceed provided that 44-<e44*444+.+ 4 :nd indteeted-ACS-f". $ a ht:1 *10u reu-47e-demonstrated, through incere flux mapping and AC&-tet:1 ' lev rate-c%.4 son, to be within 44-+et e O f-::eetu64+-
t
- etentka-@:e :n @M+-M prior to exceeding the following
'4'
..t THERMAL POWER lee i ut Z0 1.
A nominal 50% of RATED THERMAL POWER, 2.
A nominal 75% of RATED THERMAL POWER, and 3.
Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months of attaining greater than or equal to 95% of RATED THERMAL POWER.
i SURVEILLANCE REOUIREMENTS 4.2.3.1 The provisions of Specification 4.0.4 are not applicable.
F[g Qw_se 5 44c :f a a4*a4-ac5-ws:1 <!:= nte :ne e n
t be deter-a mined to be within the-+et er Of ::<ept:b!: O p: ra t4 r e f r i gu r;.te-4+ A e ' t h,
t u,5 v.,....m.
. n s.2 a t t te.,
s.e......s p. a w. u t a
-n Prior to operation above 75% of RATED TH.WAL POWER af ter each fuel a.
loa di ng, -and-1 R b.
At least once per 31 Effective Full Power'0ays, u d a
r.,
n m ec. -.4 rj', sun.. w..,.a
, c.
(,. s. u.. r
....a..st,.
.M TM inefeeted-4CS-tett+-Hew-cate-sheH-h veH f ied-u-ee-wi thica,he <
fegic t gcceptable operation of Figure 3.2-3 at least once per 12 hoverwhe n
$he most rec'TrrtL obtained value of R, obtained per Specificat14fr"472.3.2, is assu.ted to exist.
4.2.3.4 ine RCS total flow rate kitarMhall be subjected to a CHANNEL f
CALIBRATION at least once per IS e tt!!NTMie a s u r e m e n t i n s t ru.t e n t a t i o n j
8 ' N.
thall be calibrated within '
s prior to the fee mance of the calorimetric flow measure N nt.
L. 2. 3p total flow rate shall be determined by precision hea hlanc wsGe** t-e t 4 ea s t-eee e-ter-1 &- aen t e s.
-~
CCMANCHE PEAK - UNIT 1 3/4 2 10
r
,grx.x-p a78o numameoT 6
?RM 9 A O p-10 5 CPSES Technical Specificitions NRC Draft 2 Markup Section 3/4.3.1 Change 10#
Justification for Change 5110 Added note to exclude the source range neutron detector from the time response testing.
This is based on the response time of a neutron detector (in the range of milli-seconds) which becomes trivial based on the overall response time of the channel.
This makes it consistent with the other neutron detectors in the table.
5111 Add 9d note to identify that the time response incorporates the thermal well response time.
This is to compensate for the RTO response time of 5 seconds that io part of the Overpower N-16 trip function.
This is the same as the Overtemperature N-16 trip funct:on.
5112 The Steam Generator Water Level Low Coincident with Steam /Feedwater Flow Mismatch does not exist in the CPSES Design.
e a
1
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III-ettti 887Bo AITACHMENT 6 PAGE 10 0F 105 3/4.3 INSTRUMENTATION 3/4.3.1 REACTOR TRIP SYSTEM INSTRUMENTATION s
LIMITING CON 0! TION FOR OPERATION 3.3.1 As a minimum, the Reactor Trip System instru ntatio chan els and interlocks of Table 3.3-1 shall be OPERABLE with R ter T Sy em RESPONSE 6b.
TIMES :: : 5 - " Tel: 3.3-3.
w W h '. 1 W e.*w
\\,m, s.
ID 4: 0057 APPLICABILITY: As shown in Table 3.3-1.
l ACTION,:
As shown in Tad!e 3.3-1.
SURVEILLANCE REQUIREMENTS 4.3.1.1 Each Reactor Trip System instrumentation channel and interlock and the automatic trip lode shall be demonstrated OPERABLE by the performance of the Reactor Trip System Instrumentation Surveillance Requirements specified in Table 4.3-1.
I 4.3.1.2 The REACTOR TRIP SYSTEM RESPONSE TIME of each Reactor trip function shall be demonstrated to be within its limit at least once psr 18 months.
Each test shall include at least one train such that coth trains are tested at least once per 36 months and one channel per function such that all channels are tested at least once every N times 18 months where N is the total number of redundant channels in a specific Reactor trip function as shown in the "Total No. of Channels" column of Table 3.3-1.
6 l
l s
l COMANCHE PEAK - UNIT 1 3/4 3-1 L
l
+
)
i TABLE 3.3-2 188 0057 8g REACTOR TRIP SYSTEN INSTRUMENTATION RESPONSE TINES M
x
[
FUNCTIOrdt UNIT RESPONSE TIME c
x 1.
Manual Reactor Trip N.A.
g 2.
Power Range, Neutron Flux 1 0.5 ond*
3.
Power Range, Neutron Flux, g
High Positive Rate N. f..
4.
Power Range, Neutron Flux, High Negative Rate 5 0.5 second*
5.
Intermediate Range, Neutron Flux N.A.
r b 4 5 fic 6.
Source Range, Neutron Flux
$ 0.5 seconds y
7.
Overtemperature M-16
$ 7 seconds *#
Tb4 s u t 8.
Overpower M-16 1 7 seconds" Pressurizer Pressure--Low /
-< 2 seconds 9.
10.
Pressurizer Pressure-- Igh 1 2 seconds
==
11.
Pressurizer. Water, Level--High N.A.
OST 5*5 eo
Neutron / gamma detectors are u empt from response time testing. Resporse time of the neutron flux /gi signal portion of the channel shall be measured from detector output or input of first electronic component in channel
Response time includes thz thermal well response time.
D 2
21
TABLE 3.3-2 (Cont'n9ed)
O E
REACTOR TWIP SYSTEM INSTRUNENTATION RESPONSE TINES Ios,0051 ME FUNCTIONAL UNIT RESPONSE TINE 12.
Reactor Coolant Flow--Low
[
a.
Single Loop (Above P-8)
$ [1] s and z
b.
Two Loops (Above P-7 and below P-8)
$ [1 econd
-4 13.
Steam Generator Water Level--Low-tow
[2] secor.ds
-14.-Steam Cc..c. atec-htee- -
Ib4 5182.
-triel-tMeincident-with---
-Steam /Feedseter FleeMismeich N.A.
14 45.
Undervoltage - Reactor Coolant Pumps
$ [1.5] seconds 3 15 ifr. Underfrequency - Reactor Coolant Pumps 1 [0.6] second Y 4 -17.
Turbine Trip a.
Low fluid Oil Pressure M.A.
b.
Turbine Stop Valve Closure N.A.
17 18-Safety Injection Input fr SF N.A.
e - ta 11..
Reactor Trip System I erlocks N.A.
m n $ l'l 20 Reactor Trip bre ers N.A.
gg a,, d v s e," Zo2h Automatic T p and Interlock Logic N.A.
'.4 I
/
t::3
/
w 3:=
4 1
IXX-MH4 88780 ATTAteENT 6 PAGE 22 0F 1,05 TABLE 4.3-1 (Continued) ysL" TABLE NOTATIONS avhen the Reactor Trip breakers are closed and the Control Rod Drive System is c..:able of rod withdrawal.
DBelow P-6 (Intermediate Range Neutron Flux Interlock) Setpoint.
CBelow P-10 (Low Setpoint Power Range Neutr)n Flux Interlock) Setpoint.
AbovetheP-7(,[.[ower),s(tpoint.
d (1) If not perforced in previous 31 days.
(2) Comparison of calorimetric to excore power and N-16 power indication above 15% of RATED THERMAL POWER. Adjust excore channel and/or N-16 channel gains consistent with calorimetric power if absolute difference c.f the respective channel is greater than 2%.
The provisions of Specificatien 4.0.4 are not applicable to entry into MODE 2 or 1.
(3) Single point comparison of incore to excore AXIAL FLUX DIFrERENCE above 15% of RATED THERMAL POWER.
Recalibrate if the absolute ID 1: 0118 difference is creater than or equal to 3%.
For the purpose of these surveillancel"M"l requirements 71s defined as at least once per 31 days. EF Pb.
The provisions of Specification 4.0.4 are not applicable for entry into MODE 2 or /t.
l e
2.
(4) Neutron and N-16 detectors may be excluded from CHANNEL CALIBRATION.
(5) Detector plateau curves shall be oytained and evaluated.
For the Intermediate Range A 4 Power Range Mutron Flux and N-16 channels the provisions of Specjifica;.:on 4.0.4 are r.ot applicable for entry into MODE 1 or 2.
g,..
p g
.I %e AMt. F A) A BirW%C&
1080121 8
(6)
Incore-ExcoreCalibration/abovef5%ofRATEDTHERMALPOWER.For the purpose of these surveillance requi,rements "Q" is defined as at least once per 92,4ay+.
The provisions of Specification 4.0.4 are not applic-able for entry into MODE 1 or 2.
I i
EFPO.
50 (7) Each trair, shall be tested at least every 62 days on a STAGGERED TEST BASIS.
(8) The surveillance frequency and/or m s specified for these chan.1els in g,
Table 4.3-2 are more restrictive and therefore applicable.
a a
(9) Quarterly surveillance in H00ES 3, 4', <nd S shall also include verifica-tion that permissives P-6 and P-10 ar* m their required state for exist
ing plant conditions by observation of the permissive annunciator window.
Quarterly surveillance shall include verification of the Boron Oilution AlarmSetpointoflessthanorequaltofanincreaseoftwicethecount rate within a 10-minute period)'. T he 7 c.,o a t e% J Spect(;cd.en 4.o.4 a.cc not a.3p u ca Mt b
u. y
o 2.4 bots Eo do a.
o.atkor bip.
10 h 010 COMANCHE PEAK - UNIT 1 3/4 3-13 l
nLt. st76o.
Parecum44r 6 T*'8 ** A or lo s -
CPSES Technical Specifications NRC Draft 2 Markup-Section 3/4.3.2
.f Change ID#
Justification for Change 511'a The original change to have't' he sequencer's logic tested.
was' made in error since the Automatic Tester does not test all possible logic function combinations.
The Trip Actuating Device Operational-Test (TADOT) was selected-on a 62 day staggered test basis to be consistent with the slave relay testing that is the initiating signal for this test.
The exceptions to final device actuation and setpoint verification is based on design configuration which'does not allow setpoint verification without lifting leads. Also, the safeguards equipment that would have to be actuated are equi) ment of systems that are not desired to be in operation w111e the plant is 6t power.
J f
l 4
i 4
e 4
j 4
in 99tti 8A%0 AfiACHMENT 6 FAGE 31 0F 105
~
INSTRUMENTATION 3/4.3.2 ENGINEERED SAF6TY FEATURES ACTUATION SYSTEM INSTRUMENTATION i
LIMITING CON 0! TION FOR OPERATION 3.3.2 The Engineered Safety Features Actuation System (ESFAS) instrumentation channels and interlocks shown in Table 3.3-J shall be OPERABLE with their Trio 2
Setpoints set consistent with the values shown in the Trip Setpoint column of Table 3.3-f and with ESF RESPONSE TIMES ;; ;h wn in Taoit 3
3. M.- d. W.o We.W APPLICABILITY:
AsshowninTable3.3-/.
\\.mA vo.\\ce, j 2
ACTION:
10 1: 0128 I
With an ESFAS Instrumentation or Interlock Trip Setpoint trip less a.
conservative than the value shown in the Trip Satpoint column but more conservative than the value shown in the Allowable Value column of Table 3.3-f, adjust the Setpoint consistent with the Trip Setpoint value.
J b.
With an ESFAS Instrumentation or Interlock Trip Setpoint less conserva-tive than the value shown in the Allowable Value column of Table 3.3-f, either:
1.
Adjust the Setpoint consistent with the Trip Setpoint value of Table 3.3-#? and determine within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months that Equation 2.2-1 e
was satisfied for the affected channel, or a
2.
Declare the channel inoperable and anply the applicable ACTION statement requirements of Table 3.3-J2until the channel is restored to OPERABLE status with its Setpoint adjusted consistent with the Trip Setpoint value.
~
Equation 2.2-1 Z + R + 5 $ TA Where:
Z = The value from Co'umn Z of Table 3.3-for the affected chanrel, R = The "as measured" value (in percent span) of rack error for the affected channel, S = Either the "as measured" value (in percent span) of the sensor error, or the value from Column S (Sensor Error) of Table 3.3 t for the affected channel, and 3
TA = The value from Column TA (Total Allowance) of Table 3.3 for l CD-the affected channel.
With an ESFAS instrumentation channel or interlock inoperable, take c.
the ACTION shown in Table 3.3-4.
2 f
C0KANCHE PEAK - UNIT 1 3/4 3-15
III M Ba g3 ATTACHMENT 6 PAGE 41 0F 105 INSERT A g.
Not applicable if Preferred [fsite Source Breaker is open.
a, f
1
/
I-
T i
J TABLE 3.3-# (Continued)
E E
EWIINEERE. SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION IRIP SETPOINTS xn I
E SFNSOR TOTAL ERROR y
FUNCTIONAL UNIT ALLOWANCE (TA) Z (5)
TRIP SETPOINT ALLOWA8tE VALUE 7
4.
Steam Line Isolation E
a.
Manual Initiation N.A.
N.A.
N.A.
M.A.
N.A.
-.o e.
b.
Automatic Actuation Logic M.A.
N.A.
N.A.
M.A.
M.A.
and Actuation Relays j
c.
Containment Pressure--High-2 2.5 0.71 1.5 16.35 psig 16.9 psig d.
Steam Line Pressure--Low 17.3 14.81
1. 5
>605 psig*
>586 psig*
l e.
Steam Line Pressure -
8.0 0.5 0
{
Negative Rate--High
-<100 psi"*
-<111.6ps:(**
a Cb.
5.
Turbine Trip and Feedwater U
Isolation a.
Automatic Actuation Logic M.A.
N.A.
N.A.
N.A.
M.A.
and Actuation Relays b.
Steam Generator Water
7. 6, 4.3
1. 5 182.4% of 184.2% of narrow
,__ggy Level--High-High fede narrow range range instrumea.t ggg instrument span.
g g :;
Span.
Tr.,p e
-e c.
Safety Injection e Ites 1 above for all Safety Injectioytpoir.ts and [lowable S
g o
Iues.
g, -
D
t3 2=-)-
4 TABLE 4.3-2 (Continued) 8 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION f
SilRVElttANCE REQUIREMENTS M
1m TRIP ANALOG ACTUATING MODES-9 CHANNEL DEVICE MASTER SLAVE FOR WHICH
^
CHANNEL CIWINEL CHANNEL OPERATIONAL OPERATIONAL ACTUATION RELAY RELAY SURVEILLANCE FUNCTIONAL UNIT CHECK CALIBRATION TEST TEST LOGIC IEST TEST TEST IS REQUIRED
b. RW51 Level-Low-Low 5
A M
N.A.
N.A.
N.A.
N.A 1, 2, 3, 4 Coincident With s--
Safety Injection See Item 1. above for all Safety Injection Surveillance Requirements.
8. Loss of Power (6.9 kV Safeguards System Undervoltage) w a.
Preferred Offsite a
ID s: 0173 2
Source Undervoltage N. R.
R g,g.
(3 M A-M. A.
'0.Q l 1*3*q G.
w 3
j
-1)-Underveltage J
-Relay
" A.
R N A.
H-NM.
-- N. A.- - N. A.- ---- 1,- 2, -3,. 4 -
2)-Diesel-Start-iner
- N;A.
R N.A.
M -- - - - -- N. A.- - -- - N. A. - -- M: A.
172;--3 3
-3)--Source-8k e---Trip-iese N:A.
M-M:A.
N;A.
N A.
172;-3;- g gfera.jc oiM..Luc U.1,*//.N. A. -
- R - - -
4 g,a 4 g,
n A.
( 3, X) 3.
e4.A.
a. g\\,
n,g, g, g3* 4
!. -b.
Bus Undervoltage 1 og.
R, n.g.
g g)y g.g,
g,q, g 'g '
A. M r*Aed Y'It* y n.q.
R.
g.g.
>l*A.N
(
-1)S
- Diesel-Start
( 3.g
,a g '
a'3 -
M-6.
i L, 't, i
-a)- Undervol tage-M : ::
-Relay N.Ac R
- N. A c - - - - - M --- -
- - N. A.
- -- - N.A:
N: ^..
1, 2, 3, '
'" A A Ei0
-b)-T imer
- - - N. A. - - R N. A.-
---M-
-N:A.
MrA.
M-A.
172r-3; % "-. ;
W 2:=
M
TABLE 4.3-2 (Continued) oo ENGINEERED SAFETY FEsTURES ACTUATION SYSTEM INSTRtBENTATION h
$URVEILLANCE PEQUIREBENIS 9
TRIP y,
ANALOG ACTUATING N00ES' y
CHANNEL DEVICE MASTER SLAVE FOR WHICH CHANNEL CHANNEL CHANNEL OPERATIONAL OPERATIONAL ACTUATION RELAY RELAY SUstVEJLLANCE FUNCTION #.L UNI _T CHECK CALIBkATION TEST TEST LOGIC TEST TEST TEST IS REQUIRED c
h b.
Reactor Trip, P-4 M.A.
N.A.
N.A.
R N.A.
N.A.
M.A.
1, 2, 3 11.
Solid State Safeguards Sequencer (5555)
/D(t,3,4)
,J. A.
D' 5"3 a.
Safety Injection M.A.
R N.A.
-tkA-
-MflTE)-
N.A.
M.A.
1, 2, 3, 4 Sequence M U>3 4]
> \\. 4.
3 b.
Blackout Sequence N.A.
R N.A.
-M-A-
-MflTE)-
N.A.
N.A.
1, 2, 3, 4 1
O
" t =
N5{
IB 8: 0176
{Ag TABLE NOTATION h
(1) Each train shall be tested at least every 62 days on a STAGGERED TEST BASIS.
O
'g 4 2)-Perf ormed-by-Solid-State-SafetJuareis-Scgm.;ces -Atatematic !c;L.
(3) Setpoint verification is not applicable.
(4) Actuation of final devices not included.
te < n w. m - -... - a w.,
er3 ~, m pso s.c., m s m.~ ~ ay_< 6 md. a 3 <e o os ua3 as-*m*~c (
u, o+
heen 3
' T't.x.gg7eo BWRf-Hm64T 6 CPSES Technical Specifications
?nGe 6aA er tos WRC Draft 2 Markup Section 3/4.3.3 and 3/4.3.4 Chanae ID#
Justification for Chanae 5114 Normalization of the data obtained from the incore detectors is necesst.ry to properly map the core power distribution; however, it is not necessary to perform this atigt to the use of the incore detection system. The detectors a n be normalized by adjustment of the gains on each detector after passing the detectors through a common thimble thus correcting for different detector sensitivities. Normally, however, the raw data is obtained in a flux map, including passes of each detector through a common calibration thimble during the mapping.
When this data is processed by CONFORM, a ca',1bration routine compares the data, determines the detector relative sensitivity and normalizes the data.
Thus, the normalization is not done "prior to use" of the incore detectors, but rather later when the data is processed (not an on-line process).
The proposed wording is consistent with the present bases and the incore detector technical manual.
sc
VII 96tf BB rGo I
ATTACH::[NI 6 fAGE 12 # 105
,I STRUMENTATION MOV LE INCORE DETECTORS IDI0m LIMITI Ct. 40! TION FOR OPERATION
\\
3. 3. 3. 2 Th vable Incore Detection System shall be OPERABLE with:
a.
At le st 75% of the detector thimbles, b.
A minim of two detector thimbles per core quadrant, and C.
Sufficient fovable detectors, drive, and readout equipment to map these thimbr s.
APPLICABILITY: When the vable Incore Detection System is used for:
a.
Recalibration of he Excore Neutron Flux Detection System, or b.
Monitoring the QL'A0 NT POWER TILT MTIO, or N
c.
Measurement of Fg,Fg ) and F,j, g,Ig:
With the Movable Incore Detection Systeg inoperable, do not use the system for the above applicable monitoring or calibe tion functions.
The provisions of Specifications 3.0.3 and 3.0.4 are not ap icable.
b_URVEILLANCE REQUIREMENTS Ib o sing 4.3.3.2 The Novable Incore Detection System shall be monstrated OPERABLE for:
'N a. tig ngoinL etttp(ut when requiredb s within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months prior to use by nc,raelit+ng-each detect 3
--- g ss a.
Recalibration of t.he Excore Neutron Flux Detectio System, ort b.
Monitoring the QUADRANT POWER tit' RATIO, or N
y, F (Z) and F,y, cmd dd.s, r,4n c.
Measurement of F g
g Mc-e-P kbabky of4f.;
Volf S c c. u;-ye.
COMANCHE PEAK - UNIT 1 3/4 3-49
1xx-%tM usi8o ATTACHMENT 6 PAGE 93 OF 105 INSTRUMENTATION RADI0 ACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION LIMITING CON 0! TION FOR OPERATION g
($
3. 3. 3. M The radioactive gaseous effluent monitoring inetrum ntation channels showninTable3.3-W3ha11beOPERABLEwiththeirAlarm/Tri Setpoints set to ensure that the limits of Specifications 3.11.2.1 and 3.11.2 are not exceeded.
(b.
The Alarm / Trip Setpoints of thest-channels meeting Specification 3.11.2.1 shall be determined anti adjusted in accordance with the methodology and parameters in the 00CM.
APPLICABILITY:
As snown in Table 3.3-2 0 ACTION:
a.
With a radioactive gaseous effluent monitoring instrumentation channel Alarm / Trip Setpoint less conservative than required by the above specification, immediately suspend the release of radioactive gaseous effluents monitored by the affected channel, or declare the chann%) inoperable.
b.
With less than the minimum number of radioactive gaseous effluent monitoring instrumentation channels OPERABLE, take the ACTION shown in Table 3.3-12/0 Restore the inoperable instrumentation to OPERABLE status within 30 days and, if unsuccessful explain in the next Semi-annual Radioactive Effluent Release Report pursuant to Specifica-tion 6.9.1.4 why this inoperability was not corrected in a timely manner.
c.
The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS 4.3.3.10 Each radioactive gaseous effluent monitoring instrumentation chant:e1 shall be demonstrated OPERABLE by performance of the CHANNEL CHECK, SOURCE i
CHECK, CHANNEL CALIBRATION and O!GITAL CHANNEL OPERATIONAL TEST or ANALOG CHANNELOPERATIONALTESTatthefrequenciesshowninTable4.3-f.S COMANCHE PEAK - UNIT 1 3/4 3-69 l
1 1
4 m 89tt!98780 AffAWMENT6 FAGE 98 0F 105, b
s TABLE 4.3-F (Continued)
TABLE NOTATIONS
Also prior to a release from the w e
ho p sy a or containment g,
pur ng or von ng.
(1) The SIGITAL CHANNEL OPERATIONAL TF.ST shall also demonstrate that automatic isolation of this pathway and control room alarm annunciation occure, if any of the following conditions exists:
a.
Instrument indicates measured levels above the Alarm / Trip Setpoint, or b.
Circuit failure (Channel Out of Service - Loss of Power, Loss of Counts,LossofSample/ Flow,orCheckSourceFailure).
(2) The DIGITAL CHANNEL OPERATIONAL TEST shall also demonstrate that control room alarm annunciation occurs if any of the following conditions exids:
a.
Instrunient indicates measured levels above the Alarm Setpoint, or b.
Circuit failure (Channel Out of Service - Loss of Power, Loss of Counts,LossofSample/ Flow,orCheckSourceFailure).
(3) The initial CHANNEL CALIBRATION shall be performed using one or more of the reference standards certified by the National Bureau of Standards (NBS) or using standards that have been obtained from suppliers that participate in measurement assurance activities with NBS.
These standards shall permit calibrating the system over its intended range of energy and measurement range.
For subsequent CHANNEL CALIBRATION, sources that have been related to the initial calibration, reference standards certified by NBS, or standards that have bee 9 obtained from suppliers that participate in measurement assurance activities with NBS shall be used.
(4) The CHANNEL CALIRRATION shall include the use of standard gas samples g ', g g r g eil i n a.u o A m e. W W W' m M @rn 6.
000 V0l'"O percent hyd"00ea, D'l'nF8 MitPng*n-a n d__
b---Jewe-vehme-pwcont---hyd r00e a. - 6 a l a nc e a 4 t aege a a
(5) The Channel Check shall consist of visually verifying that the collection element (i.e., filter or cartridge, etc.) is in place for sampling.
CCMANCHE PEAK - UNIT 1 3/4 3-74
fir een! sta8ci AffACHMENT6 PAGE 99 0F 105 INSTRUMENTATION 3/4.3.4 TURBINE OVERSPEED PROTECTION LIMITING CONDITION FOR OPERATION 3.3.4 At least one Turbine Overspeed Protection System shall be OPERABLE.
APPICA8ILITY:
MODES 1, 2*, and 3*.
r ACTION:
Ccnktc\\
With one stop valve or one geveener valve per high pressure turbine a.
steam line inoperable and/or with one reh::t stop valve or one (b.
Conhol Meet Sterc@t valve per low pressure turbine steam line inoperable, restore the inoperable valve (s) to OPERA 8LE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , or close et least one valve in the affected steam line(s) cr isolate the turbine from the steam supply within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , b.
With the above required Turbine Overspeed Protection System otherwise inoperable, within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months isolate the turbine from the steam supply.
SURVEILLANCE REQUIREMENTS 4.3.4.1 The provisions of Specification 4.0.4 are not applicable.
4.3.4.2 The above required overspeed protection system shall be demonstrated OPERABLE:
At least once per 14 days by cycling each of.the following valves a.
through at least one complete cycle from the running position using the manual test or Automatic Tur')ine Tester (ATT):
1)
Four high pressure turbine stop valves, 2)
Four high pressure turbine control valves, Four@ Stx low pressure turbine stop valves, and Foury low prassure turbine control valves.
l b.
At least once per 14 days by testing of the two mechanical ovetspeed devices using the Automatic Turbine Tester or manual test, c.
At least once per 31 days by direct observation of the movement of each of the above valves through one complote cycle from the running position, d.
At least once per 40 months by disassembling at least one of each of ths above valves and performing a visual and surface inspection of valve seats (if applicable), disks and stems and vecifying no unaccept-able flaws.
If unacceptable flaws are found, all other valves of that type shall ba inspectad.
Not applicable it. MODES 2 and 3 with all main steam line isolation valves and associated bypass valves in the closed position.
CCMANCHE PEAK - UNIT 1 3/4 3-75
1 III 995ft 667Bo AliACHlOf6 FAGE 100 W 105 3/4.3INSTRUMENTATJON BASES 3/4.3.1 and 3rd.3.2 REACTOR TRIP SYSTEM and ENGINEERED SAFETY cEATURES ACTUATION SYSTEM INSTRUMENTATION The OPERA 8ILITY of the Reactor Trip System and the Engineered Safety Features Actuation System instrumentation <.nd interlocks ensures that: (1) the associated ACTION and/or Reactor trip will be initiated when the parameter monitored by each channel or combination thereof reaches its Setpoint (2) the specified coincidence logic and sufficient redundancy is maintained to permit a channel to be out-of-service for testing or maintenance consistent with main-taining an appropriate level of reliability of the reactor protection and engi-neered safety features instrumentation, and (3) sufficient system functional capability is available from diverse parameters.
The OPERASILITY of these systems is required to provide the overall reliability, redundancy, and diversity assumed available in the facility design for the protection and mitigation of accident and transient conditions.
The integrated operation of each of these systems is consistent with the assumptions used in the safety analyses.
The Surveillance Requirements specifled for these systems ensure that the overall system functional capability is maintained com-parable to the original design standards. The periodic surveillance tests per-formed at the minimum frequencies are sufficient to demonstrate this capability.
Specified surveillance intervals and surveillance and maintenance outage times have been determined in accordance with WCAP-10271, "Evaluation of Survei'1ance Frequencies and Out of Service Times for the Reactor Protection Instrumentation System", and supplements to that report as approved by the NRC and documented in theSER(lettertoJ.J.SheppardfrgmCecil0.Themas,datedFebruary 21, 1985).
Setpoints specified in Table 3.3-\\js Actuation System Instrumentation Tr The Engineered Safety Feature are the nominal values at which the bistables lo.
are set for each functional unit.
A Setpoint is considered to be adjusted consistent with the nominal value when the "ss me'isured" Setpoint is within the band allowed for calibratien accuracy.
3 To accommodate the instrument drift assumed to occur between oper ional tests and the accuracy to which Setpoints can be measured and calibra Allowable Values for the Setpoints have been specified in Table 3.3-/
Opera-p tion with Setpoints less conservative than the Trip Setpoint but within the Allowable Value is acceptable since an allowance has been made in the safety analysis to accommodate this error.
An optional provision has been inoluded for datermining the OPERA 81LITY of a channel when its Trip Setpoint is found to exceed the Allowable Value.
The methodolcgy of this option utilizes the "as measured" deviation from the specified calibration point for rack and sensor components in conjunction with a statistical combination of the other uncertainties of the instrumentation to measure the process variable and the uncertainties in calibrating the instrumentation.
In Equation -H-i, M-I h b-Z + R + 5 < TA, the intoractive effects of the errors in the rack and tre sensor, ani$ the "as measu!;ed" values of the errors are considered.
Z, as specified in Table 3.3 Win percent span, is the statistical summation ja of errors assumed in the analysis excluding those associated with the sensor and rack drift and the accuracy of their measurement.
TA Jr Total Allowance
/
is the difference, in percent span, R or Rack Error is the "as measured" deviation, in the percent span, for the af fected channel from the specified Trip Setpoint.
S or Sensor Error is eitner the "as measured" deviation of COMANCHE PEAK - UNIT 1 8 3/4 3-1
TII 809t 881Bc>
Ai?ACHMEW 6 PAGE 102 CF 105 INSTRUMENTATION BASES REACTOR TRIP SYSTEM and ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION (Continued)
L.1 3
the sensor from its calibration point or the value specified in able3.3-/,
O.
in percent span, from the analysis assumptions.
Use of Equation 1 allows for a sensor draft factor, an increased rack drif t factor, and provides a threshold value for REPORTABLE EVENTS.
The methodology to derive the Trip Setpoints is based upon combining all of the uncertainties in the channels.
Inherent to the determination of the Trip Setpoints are the magnitudes of these channel uncertainties.
Sensor and rack instrumentation utilized in these channels are expected to be capable of operating within the allowances of these uncertainty magnitudes.
Rack drift in excess of the Allowable Value exhibits the behavior that the rack has not met its allowance.
Being that there is a small statistical chance that this will happen, an infrequent excessive drift is expected.
Rack or sensor drif t, in excess of the allowance that is more than occasional, may be indicative of more serious problems and should warrant further investigation.
The measurement of response time at the specified frequencies provides assuran e that the Reactor trip and the Engineered Safety Features actuation associated with each channel is completed within the time limit assumed in the safety analyses.
No credit was taken in the analyses for those channels with response times indicated as not applicable.
Response time may be demonstrated by any series of sequential, overlapping, or total channel test measurements provie.'ed that such tests demonstrate the total channel response time as defined.
Sensor response time verification may be demonstrated by either:
(1) in place, onsite, or offsite test measurements, or (2) utilizing replacement sensors with certified response time.
The Engineered Safety Feetures Actuation System senses selected pl, ant parameters and determines whether or not predetermined limits are being l
exceeded.
If they are, the signals are combined into logic matrices sensitive to combinations indicative of various accidents events, and transients.
Once the required logic combination is completed, the system sends actuation signals to those Engineered Safety Features components whose aggregate function best serves the requirements ef the condition.
As an example, the following actions may be initiated by the Engineered Safety Features Actuation System to mitigate the consequences of a steam line break or loss-of-coolant accident.
(1) ECCS pumps start and automatic valves position, (2) 8 tactor trip (3) feed water isolation, (4) startup of the emergency diesel generators. (5) containment
)
rpray pumps start and automatic walves position (6) containment isolation, (7) steam line isolation, (d) turbir,e trip, (9) auxiliary feedwater pumps start and automatic valves position, (10) station service water pumps s' art and i
automatic valves position, (11) Control Room Emergency Recirculation starts, l
and (12) essential ventilation systems (safety chilled water, electrical area l
fans, primary plant ventilation ESF exhaust fans, battery room s=haust fans, I
and UPS ventila-ion) start.
(
CCMANCHE PEAK - UNIT 1 B 3/4 3'2
10-0%tt GM80 ATTACMENT 6 PAGE1040F1%
i e
INSTRUMENTATION ID I: 0/66 BASES I
REMOTE SHUTDOWN SYSTEM (Continued) g gy Awm Am(e,. SAL,s %L notprecludeachievingsafe[ shutdown. emote [hutdown nsures that a fire will The OPERA 8ILITY of the The mote utdown Gye4+m instrumenta-tion, control; :nd ;;;- r circ.its and transfer switches necepsary to eliminate effects of the fire and allow operation of instrumentation (Yontrols;nd ;:1:r cir;.its requ', rad to achieve and maintain a safe shutdown condition are inde-pendent of areas where a fire could damage systems normally used to shut down the reactor.
This capability is consistent with Ger. oral Design Criterion 3 of 10 CFR 50.
3/4. 3. 3. 6r5 ACCIDENT MONITORING INSTRUMENTATION~
The OPERA 8ILITY of the accident monitoririg instrumentation ensures that sufficient information is available on selected plant parametars for which pre-planned manually controlled operator actions are required to accomplish safety functions for recovery from Design Basis Accidents, as defined by the plant safety analysis.
This capability meets the intent of the recommendations of Regulatory Guide 1.97, Revision 2, "Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant Conditions During and Following an Accident," December 1980 and those requirements of NUREG-0737, "Clarification of TMI Action Plan Requirements," Novemeer 1980 that apply to CPSES.
m_
___a.......,,--.. __,... s e.... e........ m...
...i. m.
h ag; O 5 5 iIi h 5 5 bA5E U !5 d d: l 5Nr:55 *5b Si lebEE-h by t^k i"'--
th;; St: : h::nd:ry Cee!:%-Ani!Mi'ity fu etica !! coat i
+^^t "4+" Ac W "
^ ion s w u ' i..... i. II.C.1.2 ef NU"EC-0727 f;r ";;tingh: :: " P00 U r ! !"' W * *
w;;;t;r;.
gg cm The soecific calibration provisions for the Containment Radiation (High Range) Monitor are in accordance with the provisions of NUREG-0737, Ites II.F.1.
3/4.3.3.[CHLORINEDETECTIONSYSTEMS The OPERABILITY of the chlorine Detection Systems ensures that sufficient capability is available to promotly detect and initiate protective action in the event of an accidental chlorine release.
This capability is required to protect control room personnel and is consistent with the recommendations of Regulatory Guide 1.95, Revision 1, "Protection of Nuclear Power Plant Control room Operators Against an Accidental Chlorine Release," January 1977.
This capability will not be required if the quantity of chlorine gas stored onsite is small (< 20 lbs.) and utilized for laboratory and calibration purposes.
Thisapplicabilityisconsistentwithtneexclusionsandrecommendationsof Regulatory Guide 1.95, Revisinn 1, "Protection of Nuclear Plant Control Room Operators Against an Accidental Chlorine Release,' January 1977.
4bl.3.3.9' LOOSE PART DETECit0N SYSTEM mN TYMERASILITY of the Loose-Part Detection System ensures that sufficient capability is av'a'Mehl(to detect loose metallic parts in the Reactor System and avoid or mitigate damag+.4o Reactor System components.
The allowable out of-service times and surveTTTanca tequirements are consisteat with the recommendations of Regulatory Guide 1.1.T7d"Loose-Part Detection Program for the Primary System of Light-Water-Cooled Reactli." Ma 1981.
l CCMANCHE PEAK - UNIT 1 8 3/4 3 5
III-9Mtf bbl &o ATTACHMENT 1 PAGE 6 Of 66 CPSES Technical Specifications NRC Oraft 2 Markup s
Section 3/4.4 (cont.)
Actions) and Surveillance Requirements has been made to appropriately implement this change.
This change is similar to that licensed at Millstone, Byron, Seabrook, Callaway, Vogtle and Wolf Creek.
0470 See 10# 0223 0475 See 10# 0233 0478 See 10# 0240 0480 See IDF 0242 0481 See !Of 0237 0906 This Action Statement is revised to allow following the appropriate Action Statement for the plant condition being addressed.
This change in no-way lessens the protection afforded by any of the present Action Statements. Based on 3.0.2, without this change once the Action Statement is entered it cannot be deviated from until the LC0 is met.
This is considered to be overly restrictive in this situation.
s 0937 Thi Technical Specification is being relocated to the echnical Specification Improvement Pro ram.
TV
$ 7~ N "M Y o CP$E Electri believes the inclusion of-this Speci ication is unnecessa and the information would be more
-((p
Men appropriate addressed in the CPSES Technical Specification e rovement Program.
Relocation of this Specification is consistent with the guidance provided in 'the NRC's Interim Policy Statement (52FR3788), February 6,'1987, and the recommendations of the Westinghouse Owners Gr MERITS Program.
Priority is given to the reloca on of this Specification since the detailed information is t used by the Licensed l
Operator, and requires no imediate tion from the Licensed Operator if the Action Statem t is applied.
The information currently in this Specificat is more appropriately maintained in a document sub to TV Electric administrative control and 10CFR50.5 view under the CPSES Technical Specification improvem Program.
IIX 4HW 8EnBo AffetW O i 7 FAGE 7 0F 66 CPSES Technical Specifications NRC Draft 2 Markup Section 3/4.4 Chanae_108 Justification B e Chance 0937 (cont.)
hi g
rris, 0938 The Valves 8705A/B should be deleted from Table 3.4-1.
Table 3.4-1 contains a list of RCS pressure isolation valves for which operational leakage shall be limited to 1 gpm during MODES 1,2,3 and 4.
Tbc NRC's intent in this specification is to limit intersystem loss of coolant from the RCS to major depressurized subsystems.
In particular, the concept was to limit leakage past the SAFETY INJECTION SYSTEM check valvos in addition to the RHR suction isolation valves.
Valves 8705A/B are 3/4-inch check valves which provide a means of pressure relief from the volume of trapped water between the RHR suction isolation valves.
Valves 8705A/B are not classified as Reactor Coolant System pressure isolation valves despite the fact they are in parallel with the innermost RHR suction isolation valves.
This is due to the fact that valves 8705A/B are located in a flowpaths which have orifices to prevent RCS leakage in excess of the makeup system.
If this valve failed to isolate RCS pressure, the potential mass flow into the RHR system could,be handled by the RHR suction reliefs.
O W
T4x - %7 E O ATT A CHmcc 7 CPSES Technical Specifications Pgc 7 A e r e.g NRC Draft 2 Markup Section 3/4.4 Chanae IC#
Justification for Chance 0937 See justification fo.- Change 10# 5104 in Section 3/4.0.
This supersedes the justification for ID# 0937 in TV Electric Letter TXX-88512.
It is requested that the subsequent specifications not be renumbered based on this change but that the index list Section 3/4.4.5 as not i
used, b
5115 Relocates Table 3.4-1 based on a draft of the genoric letter thct allows this relocation.
This change was discussed in the meetir.g with the NRC on October 19, 1988.
5116 This addition to the Bases is made to clarify the notch in the cooldown curve for CPSES based on the new 10CFR50 Appendix G rule.
IXI-99MtGB7e,o AITACHMENT 7 PAGE 20 0F 66 REACTOR COOLANT SYSTEM 4.4.5 STEAM GEF'ERATORS LIMI NG CON 0! TION FOR OPERATION 3.4.5 Eac steam generator shall ba OPERABLE.
APPLICA81 LIT HODES 1, 2, 3 and 4 ACTION:
With one or core 5 am generators inoperable, restore the inoperable generator (s) to OPERABLE status ior to increasing T,yg above 200*F.
[N L( f6 Ltd f D OC9 3 7 _
g_
SURVE!LLANCE REQUIREMENT 1i
( KELecnrG t b u eq 3 7 T _S9NklLL(We_& R GetfMNTS) 4.4.5.0 Each steam generato shall be demonstrated 0F IRA 8LE by performance of th's following augmented inserv e inspection program and the requirements of Specification 4.0.5.
4.4.5.1 Steam Generator Sample Se etion and Inspection - Each sceam generator shall be determined OPERABLE dJring~ nutdown oy selecting and inspecting at
(
least the minimum number of steam gen rators specified in Table 4.4-1.
4.4.5.2 Steam Generator Tube Samole se getion and Inspection - The steam generator tuce minimum sample size, inspe tion result classification, and the corresponding action required shall be as ecified in Table 4.4-2.
The inservice inspection of steam generator tube shall be performed at the fre-quencies specified in Specification 4.4.5.3 a the inspected tubes shall be verified acceptable per the acceptance criteria of Specification 4.4.5.4.
The tubes selected for each inservice inspection sha 1 include at least 3% of all the expanded tubes and at least 3% of the remaini number of tubes in all steam generators; the tubes selected for these insp ctions shall be selected on a random basis except:
I a.
Where experience in similar plants with simil wr.ter chemistry l
indicates critical areas to be inspected, then at least 50% of the l
tubes inspected shall be from these critical are'as; I
The first sample of tubes selected for each inserv\\
b.
ice inspection (subsequent to the preservice inspection) of each steam generator shall include:
N N
\\
\\
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s COMANCHE PEAK - UNIT 1 3/4 4-13
TII 966tf SB % o AllA$ MENT 7 PAGE 27 0F M REACTOR COOLANT SYSTEM S
3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE LEAKAGE DETECTION SYSTEMS LIMITING CONDITION FOR OPERATION 3.4.6.1 The following Reactor Coolant System Leakage Detection.lystems shall be OPERABLE:
a.
The Containment Atmosphere Particulate Radioactivity Monitoring
System, b.
The Coc.tainment Sump Level and Flow Monitoring System, and c.
Either the containment air cooler condensate flow rate or the Con-tainment Atmosphere Gaseous Radioactivity Monitoring System.
APPLICA8ILITY: MODES 1, 2, 3, and 4.
ACTION:
With only two of the above reouired Leakage Detection Systems OPERABLE, operation may continue for up to 30 days provided grab samples of the contain-ment atmosphere are obtained and analyzed at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months when the required Gaseous or Particulate Radioactive Monitoring System is inoperable; otherwise, be in at least HOT STAN08Y within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTOOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .
The provisions of Specification 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS 4.4.6.1 The Leakage Detection Systems shall be demonstrated OPERABLE by:
a.
Containment Atmosphere Gaseous and Particulate Monitoring Systems-perfnrJance of CHANNEL CHECX, CHANNEL CALIBRATION, and O!GITAL CHAN'.dL OP,ERATIC.1AL TEST at the frequencies :p::i:
', T:ti
!. 2 - >x op.
pw.dsd u S,. u :fr u.t w 4.2.i.s, b.
Containment Sump Level and Flow Monitoring System performance of CHANNEL CALIBRATION at least once per 18 months, and c.
Containmefit Air Cooler Condensato Flow Rate Monitoring System -
performance of CHANNEL CALIBRATION at least once per 18 months.
COMANCHE PEAK - UNIT 1 3/4 4-20
I1X-f4%F &M 80 lifiACHMENT1 PAGE 28 0F 66 REACTOR COOLANT SYSTEM OPERATIONAL LEAKAGE LIMITING CON 0! TION FOR OPERATION 3.4.6.2 Reactor Coolant System leakage shall be limited to:
a.
No PRESSURE BOUNDARY LEAKAGE, b.
1 GPM UNIDENTIFIED LEAKAGE, c.
1 GPM total reactor-to-secondary leakage through all steam generators not isolated from the Reactor Coolant System and 500 gallons per day through any one steam generator not isolated from the teactor Coolant System, d.
10 GPM IDENTIFIED LEAKAGE from the Reactor Coolant System, e.
40 GPM CONTROLLED '.EAKAGE at a Reactor Coolant System pressure of 2235 2 20 psig, and f.
0.5 GPM leakage per nominal inch of valve sin up to a maximum of 10 i: 0230 5 GPM at a Reactor Coolant System pressure of 2235 2 20 psig from any Reactor Coolant System Pressure Isolation Valve.:p:;if f:d '-
-T:ble 3.'-ic
5"6 APPLICABILITY:
MODES 1, 2, 3, and 4.
ACTION:
a.
With any PRESSURE BOUNDARY LEAKAGE, be in at least Hni STANOBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUT 00WN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .
b.
With any Reactor Coclant System leakage greater than any one of the above limits, excluding PRESSURE BOUNDARY LEAKAGE and leakage from Reactor Coolant System Pressure Isolation Valves, reduce the leakage rate to within limits within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .
i c.
With any Reactor Coolant System Pressure Isolation Valve leakage greater than the above limit, isolate the high pressure portion of the affected system from the low pressure portion within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months by use of at least two closed manual or deactivated automatic valves, or be in at least HOT STANOBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLO SHUTOOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .
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in-406H B&%o AfiA;HMENT 7 PAGE 29 0F 66
'=
m REACTOR COOLANT SYSTEM s
OPERATIONAL LEAKAGE SURVEILLANCF REQUIREMENTS 4.4.6.2.1 Reactor Coolant System leakages shall be demonstrated to be within each of the above limits by:
Monitoring the Reactor Coolant System Leikage Detection System a.
required by Specification 3.4.6.1 at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months ; b.
Measurement of the CONTROLLED LEAKAGE to the reactor co' ant pump sealt when the Reactor Coolant System pressure is 2L.' 4 20 psig at ID?: 5233 least once per M-d+y: with-th: ::ilating -valve fe!!y
^^=a
+The
~
provisions of 3pecification 4.0.4 are not applicable for entry into MODE 3 or 4; Performance-of-a-Reectek a x g r A ID I: 0234 oelent-System-**t+e-4=.terj balance et c.
loest 0 ce per 72 heure; and d.
Monitoring the Reactor Head Flange Leskoff System at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .
Tb 4 505 4.4.6.2.2 Each Reactor Coolant System Pressure Isolation Valve :;;;i'i:d '-
T:b1: 3.F1 shall be demonstrated OPERA 8LE by verifying leakage to be within its limit:
a.
At least once per 18 months, b.
PriortoenteringMODE2whenevertheplanthisbeeninCOLD SHUTDOWN for 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or more and if leakage testing has not been performed in the previous 9 months, except for valves 8701A, 8701B, 8702A, and 87028.
c.
Prior to returning the valve to service following maintenance, repair or replacement work on the valve, and d.
Following check valve actuation due to flo'w through the valve, e.
As outlinwd in the ASHE Code,Section XI, paragraph IW-3427(b).
The provisions of Specification 4.0.4 are not applicable for entry into MODE 3 or 4.
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CCMANCHE PEAK - UNIT 1 3/4 4-22
TII9Mtt S876o ATTACMENT 7 PM E 31 of 4 s,
TABLE 3.4-1
'N REACTOR COOLANT SYSTEM PRESSURE ISOLATION VALVES s
VALVE N ER FUNCTION 8948 A, B, C, Accumulator Tank Discharge 8956 A, B, C, D.
Accumulator Tank Discharge 8905 A, B, C, O SI Hot Leg Injection 8949 A, B, C, O SIHotLegInjection 8818 A, B, C, O Cold Leg Injection B819 A, B, C, O SI d Leg Injection 8701 A, B RHR Suc on Isolation
((g2A,g
[Hg jugp j g_ Is g } a} j gn,._,.,.._,,
Hi0918 8841 A, B RHR Hot Leg Inje ion i
8815 CCP Cold Leg In;'ect 8900 A, 8, C, O CCP Cold Leg In;ection
r. b.*r $11 T' ELOCArq I
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COMANCHE PEAK - UNIT 1 3/4 4 23 i
i
TXX WHe 8878o AtiACMENT ?
PAGE 34 0F %
REACTOR COOLANT SYSTEM s
3/4.4.8 SPECIFIC ACTIVITY LIMITING CONCITION FOR OPERATION 3.4.8 The specific activity of the reactor coolant shall be limited to:
a.
Less than or aqual to 1 mi:rocurie per gram COSE EQUIVALENT I-131, and b.
Less than or equal to 100 4 microcuries per gram of gross raoicactivity.
APPLICABILITY: MODES 1, 2, 3, 4, and S.
ACTION:
MODES 1, 2 and 3*:
a.
With the specific activity of the reactor coolant greater than 1 microcurie per gram DOSE EQUIVALENT I-131 for more than 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months during one continuous time interval or exceeding the limit line shown on Figure 3.4-1, be in at least HOT STANOBY with T less tha.1 500'F within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ; and avg b.
With the specific activity of the reactor coolant greater than 1
1004 microCuries per gram, be in at least HOT STANOBY with T less than 500*F within 6 hcurs.
avg r
MODES 1, 2, 3, 4, and 5:
With the specific activity of the reactor coolant greater than 1 microcurie per gram OOSE EQUIVALENT I-131 or greater than 1004 micro-Curies per gram, perform the samplins and analysis requirements of item 4.a) of Table 4.
until the specific activit of the reactor coolant is l0 restored t within its limits.
l 3
SURVE!LLANCE REQUIREMENTS 4.4.8 The specific activity of the reactor coolant shall be determined to be within the limits by performance of the sampling and analysis piogram of Table 4.4-/.
l0.
With T,yg greater than or equal to 500*F.
COMANCHE PEAK - WIT 1 3/4 4-26 l
TII M6H 88780 ATTACMENT 7 FAf.E $1 0F 64 D Msa REACTOR COOLANT SYSTEM s
8ASES 3/4.4.4 RELIEF VALVES I
The power-operated relief valves (PORVs) and steam bubble function to l
relieve RUS pressure during all design transients up to and including the design step load decrease with steam damp. Operation of the PORVs minimizas i
t'he undesirable Eining of the spring-loaded pressurizer Code safety valves.
Each PORV bas a remotely operated block valve to provide a positive shutoff capability should a relief valve become inoperable, h4.4.5 STEA" GENERATORS he Survuillance Requirements for inspection of the steam generator tubes ensure hat the structural integrity of this portion of the RCS will be main-tained.
he program for inservice inspection of steam generator tubes is based on a modif ation of Regulato y Guide 1.83, Revision 1.
Inservice inspection c' steam gen ato= tubing is essential in order to maintair' surveillance of the conditions f the tubes in the event that there is evidence of mechanical dange or progre ive degradation due to design, manufacturing errors, or it. service conditio that Iwad to corrosion.
Inservice inspection of steam generator tubing als rovides a means of characterizing the nature and cause of any tube degradatio so that corrective measures can be taken.
l Seir.ted tubes in the reheater section of each 04 and 05 steam generator f
have bm ~dified 1.o correc the tube vibration degradation phenomenon experi-enced N :
ain Westinghouse eam generators.
The r.odification,,onsisted of 2
expand y tnese tubes in the vic ity of the support plates and is designed to limit t.1e amplitude of vibration.
hese expanded tubes are subject to a special inspection whenever the stea enerators are opened for inservice eddy current testing.
]'
The M ant is expected to be operated a manner such that the secondary coolant will be maintained within those ches try limits found to result in l
negligible corrosion of the steam geneiator tu s.
If the recondary coolant i
chemistry is not maintained within these limits, ocalized corrosion may likely 3
result in stress corrosion cracking.
The extent o cracking dJring plant operation would be lirited by the limitation of stea entrator tube laakage between the Reactor Coolant System and the Secondary C lant System (reactor-te-secondary leaksge = 500 gallons per day per steam gen ator and a total laakage of 1 GPM te # steam generators).
Cracks having reactor to-secondary leaka'
'1an thO limit durirg operation will ave an adequate j
margir of safet rr :sstan<t < a 'oads imposed during normal erstion and by postulated sccir'.
Operat
'O ts have demonstrated that r c to r-t o-secondary h akss
+
'N
,e
~.ay per steam generator can r dily be detected by rat:.,
- sam generator blowdown.
Leakage n excess 1
of this limit t' i'
i' 'down and an unscneduled inspectio r
daring which ttn 1.
.e located and plugged.
j RELouTD od 04 27 CCMANChe IAK - UNIT 1 3 3/4 4-3
III 8M M 36 7 3 e ATTACHMENT 1 PAGE 52 0F 66 REACTOR COOLANT SYSTEM 8ASES L..
s s
STIAM GENERATORS (Continued) ge-type defects are 'Jnlikely with proper chemistry t x.condary ant.
However, even if a defect should develop in servi will be found reatment of the ng scheduled inservice steam gene Plugging will be re ce, it plugging ifmit of 40% o ons.
tube inspections of operatine tube nominal wall thickness.
reliably detect degradation thatlants have demonstrated the capability to thickness.
pehetrated 20% of the original tube =0 1 Whenever the results of any steam gene fall into Category C 3, these results will be Special Report pursuant to 10 CFR 50.72 within 4 hor tubing r
ted to the Commission in a and pursuant to S plant operation. pecification 6.9.2 within 30 days and ou from initial discovery casa basis and may result in a requirement for analysis tions, tes*,s, additional eddy-
. laboraton on a case-by-Specifications, if necessary. current inspection, and revision of the examina-3/4.4.6 hnical
_ REACTOR COOLANT SYSTEM LEAKAGE
_3/4.4.6.1 RE#
LEAKAGE DETECTION SYSTEMS provided to monitor and detect leakage from the rT I
boundary.
s specification are These Detection Systems are consistent with the recoea Regulator Systems,"y Guide 1.45, "Reactor Coolant Pressure Bou May 1973.
mmendations of for restoration since two di"arse and redundant R
~
, 30 days are permitted remain OPERA 8LE particulata monitoring system, grab s ment gaseous or.
as a backup to the single remaining atmospheric monit a n-erformed 3/4.4.6.2 i
ofERATIONAL LEAKAGE oring system.
be indicative of an impending gross failure of thPkES ceptable since it may the presence of any PRESSURE 80VN9ARY LEAXAGE r e pressure boundary.
Therefore, placed in COLD SHUTDOWN.
equires the unit to be promptly expected from the RCS, the umidentified pertio of leakage is to a threshold value of less than 1 gpm j
low tc ensure early detectior. of additional leakageThis threshold ge can be reduced erators not isolated from the RCS ensures that tube leakage will be limited to a small fraction of 10 CFRm for a line values in the event of either a steam generator t b contribution from the 1
4 *4 100 dose guide-break.
The 1 gem limit is consistent with the assumptionsu e rupture or s of these accidents.
steam generator tube integrity is maintained in the event rupture or under LOCA conditions.
es that a main steam line CCMANCHE PEAK - UNIT 1 g 3/4 4,4
TH0$Hf8Ingt AliACH W I ?
l PAGE 53 0F M REACTORCOOLANTSYJTEM g
i s
5I 8ASES OPERATIONAL LEAKAGE (Continued)
The 10 gpa IDENTIFIED LEAKAGE limitation provides allowance for a limited amount of leakage from known sources whose presence will not interfere with l
the detection of UNIDENTIFIED LEAKAGE by the Leakage Detection Systems, I
i The CONTROLLED LEAKAGE limitation restricts operation wha.. the total flow i
supplied to the reactor coolant pump seals exceeds 4C gpm with the modulating WV id valva.in the supply line fully ope %at a nominal RCS pressure of 2235 psig.
This limitation ensures that in the event of a LOCA, the safety injection flaw willnotbelessthanassumedythesafetyanalyses.
. ID I: 005 l
ti sts "t-I i
ensureearlydet[ectionofpossiblein-seriesvalvefailure. rom any RCS press lGb.
The leakage 4
It is apparent i
that when pressure isolation is provided by two in-series valves and when failure of one valve in the pair can go undetected for a shstantial length of time, verification of valve integrity is required.
Since t*4se valves isre 1
important in,mreventing overpressurization and rupture of the Et':S low pressure piping which coJ14 result in a LOCA, these valves should be tested periojically i
l to ensure low probability of gross failure.
i 4
The Surveillance Requirements for RCS pressure isolation valves provide f
4 added assurance of valve integrity thereby reducing the probability of { aoss j
valve failure and consequent intersystem LOCA, Leakage from the RCS pressure l
isolation valve is IDENTIFIED LEAKAGE and will be considered as a portion of i
the allowed limit.
j r
i 3/4.4.7 CHEMISTRY f
i The limitations on Reactor Coolant System chemistry ensure that corrosion f
I of the Reactor Coolant System is minimized and reduces the potential for i
i Reactor Coolant System leakage or failure due to stress corrosion. Maintaining
{
the chemistry within the Steady-State Limits provides adequate corrosion protection to ensure the structur31 integrity of the R9 actor Coolant System l
l over the life of the plant.
The associated effects of exceeding the oxygen, chloride, and fluoride limits are time and temperature dependent.
Corrosion l
studies show that operation may be continued with contaminant conce-tration j
j levels in excess of the Steady-State Limits, up to the Transient Limits, for 3
the specified limited time intervals without having a significant effect on i
the struct'>ral integrity of the Reactor Coolant System.
The time interval permitting continued operation within th>s restrictions of the Transient Limits i
provides time for takirg corrective act'oas to restere the contaminant concen-
[
4 trations to within the Steady-State Limits.
j i
The Surveillance Requirements provide adequate assurance that concentrations I
l in excess of the limits will be detected in sufficient tima to take corrective j
action.
ror~a t le % fim, a.ad cLah ed bm o'"AC #C #dr*lk d h NCN N 4C f
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C?MANCHE PEAK - UNIT 1 8 3/4 4-5 j
i i
TII40Hf 86'760 AliACm(Ni1 PAG ( 54 W 66 REACTOR COOLANT SYSTEM 8ASES 3/4.4.8 SPECIFIC ACTIVITY The limitations on the specific activity of the reactor coolant ensure
hat the resulting 2-hour doses at the EXCLUSION AREA BOUNDARY (EAB) will not e tceed an appropriately small fraction of 10 CFR 100 dose guideline values i
fo)1owing a steam generator tube rupture accident in conjunction with an assumed steady-state reactor-to-secondary steam generator leakage rate of 1 gpa.
The values for the limits on specific activity represent limits baseo upon a parametric evaluation by the NRC of typical site locations.
These values are conservative in that specific site parameters of the CPSES site.
l such as EA8 location and meteorological ccnditions, were not considered in this evaluation.
The ACTION statement permitting POWER OPERATION to continue for limited time periods with the reactor coolant's specific act'vity greater thar.
I 1 microcurh/ gram DOSE EQUIVALENT I-131, but within the allowable limit shown on Figure 3.4-1, accommodates possible iodine spiking phenomenon which may occur following changes in THERMAL POWER.
The sample analysis for determining the gross specific activity and I can exclude the radioiodines because of the low reactor coolant limit of 1 microcurie /
grae DOSE E WIVALENT I-131, and hecause, if the limit is exceeded, the radio-iodine level is to be determined every 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .
If the gross specific activity level and radiciodine level in the reactor coolant were at their limits, the radiofodine contribution would be approximately 1%.
In a release of reactor coolant with a typical mixture of radioactivity, the actual radioiodine contri-bution would probably bg about 20%.
The exclusion of radionuclides with half-lives less than 10 minutes from these determinations has been made for several reasons.
The first consideration is the difficulty to identify short-lived radionuclides in a sample that requires a significant time to collect, i
j transport, and analyze.
The second consideration is the predictable delay time between the postulated releast of radioactivity from the reactor coolant to its I
release to the environment and transport to the EAB, which is relatable to at 1eart 30 minutes decay time.
The choice of 10 minutes for the half-life cutoff was made because of the ncclear characteristics of the typical reactor coolart l
radioactivity. The radionuclides in the typical reactor coolant have half-i lives of less than 4 minutes or half-lives of greater than 14 minutes, which allows a distinction between the radionuclides above and below a half-life l
of 10 minutes.
For ths q reasons the radioauctides that are excluded frca consideration are expe%ed to decay to very low levels before they could be I
transported from the reactor coolant to the EAB under any accident condition.
The activity levels allowed by Figure 3.4-1 increase the 2-hour thyroid dose at the EA8 by a factor of up to 20 following a postulated steam generator tube rupture.
Therefore, operation with specific activity levels exceeding the l
limits of Specification 3.4.8 requires additional sampling per Table 4.4-/& 3 i
l reporting of operational and sample information in the Annual Report pursuant to l
Specification 6.9.1.4.
This is in confortrince with Generic Letter 85-19 to i
allow NRC evaluation.
COMANCHE PEAK - UNIT 1 8 3/4 4-6 i
l TII G444 88180 l
ATTACMfMT 1 f
PAGE 64 0F 66 REACTOR COOLANT SYSTEM s
BASES PRESSURE / TEMPERATURE LIMIT 3 (Continued) do not offset each other and the pressure-temperature curve based on steady-state conditions no longer represents a lower bound of all similar curves for finite heatup rates when the 1/4T flaw is considered Therefore, both cases have to be analyzed in order to assure that at any coolant temperature the lower value af the allowable pressure calculated for steady-state and finite heatup rates is obtained.
The second portion of the heatup analysis concerns the calculation of pressure-temperature limitations for the case in which a 1/4T deep outside surface flaw is assumed.
Unlike the situation at the vessel inside surface, the thermal gradients established at the outside surface during heatup produce stresses whir.h are tensile in nature and thus tend to reinforce any pressure stresses present.
These thermal stresses, of course, are darendent on both the rate of heatup and the time (or coolant temperature) along the heatup ramp.
Furthermore, since the thermal stresses at the outside are tensile and increase with increasing heatup rate, a lower bound curve cannot be defined.
Rat.ier, each heatup rate of interest must be analyzed on an individcal basis.
Following the generation of pressure-temperature curves for both the steady-state and finite heatup rate situations, the final limit curves are produced as follows.
A composite curve is constructed based on a poir.L-by-point comparison of the steady-stato and finite heatup rate data. At any given temperature, the allcwable pressure is taken to be the lesser of the three values taken from the curves under consideration.
The use of the composite curve is necessary to set conservative heatup limitations because it is possible for conditions to e'xist such that over the course Of the heatup ramp the controlling condition switches from the inside to the outside and the pressure limit must at all times be based on analysis of the most critical criterion.
rbeVHb M5Etr[
Finally, the composite curves for the heatup rate data and the cooldown rate data are arijusted for possible errors in the pressure and teft.perature sensing instruments by the values indicated on the respective curves.
Although the pressurizer operates in tempersttre ranges above those for wnich there is reason for concern of nonductile failure, operating limits are provided to assure competibility of operation with the fatigue analysis performed in accordance with the ASME Coce requiraments.
!DI:0460 N
LOW TEMPERATURE OVERPRESSURE PROTECTION 2
The OPERAB!t,ITY of two PORVsfor an RCS vent opening of at least 2.98 square inches ensures that the RCS will be protected from pressure transients which could exceed the limits of 10 CFR 50 Appendix G when one or more of the RCS cold legs are less than or equal to 350*F.
Either PORVess adequate relieving capa-bility tc protect the RCS from overpressurization when the transient is limitec to either:
(1) the start of an idle RCP with the secondary water temperature of I
the Steam generator less than or equal to 50*F above the RCS cold leg temoere tures, or (2) the start of e'iarging pumps and their injection into a water-solid g
RCS.
y CU hNCHi PEAK - UNIT 1 0 3/4 4-15 W.\\ set &
ID I: O N
]
m tx-agigo RTrQCo m(uT-7 7eG6 64 A os 66 INSERT FOR PAGE B 3/4 4-15 The new 10CFR50 Appendix G rule addresses the metal temperature of the closure head flange and vessel flange regions.
This rule stat?3 that the minimum metal temperature of the closure flange region should be at least 120 degrecs-F higher than the limiting RTNDT for these regions when the pressure exceeds 20 percent of the preservice hydrostatic test pressure (621 psig for Westinghouse plants).
For Comanche Peak Unit 1, the minimum temperature of the closure flange and ver.sel flange regions is 160 degrees-F since the
.imiting RTNOT is 40 degrees-F (see Table B 3/4.4-1).
The Comanche Peak Unit I cooldown curves shown in Figure 3.4-3 are impacted by this new rule, and therefore the "notch" in the cooldown curves.
1 l
I i
1 1
i l
l i
l
III49ttt 88780 NGE 6 # 15 I
},f
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EMERGENCY CORE COOLING SYSTEMS D'
3/4.5.2 ECCS SUBSYSTEMS - T,y, > 46 350'F LIMITING CONDITION FOR OPERATION 3.5.2 Two independent Emergency Core Cooling System (ECCS) subsystems shall be OPERA 8LE with each subsystem comprised of:
a.
One OPERA 8LE centrifugal charging pump, b.
One OPERA 8LE Safety Injection rump, c.
One OPERABLE RHR heat exchanger, d.
One OPERA 8LE RHR pump, and e.
An 0PERA8LE flow path capable of taking suction from the refueling water storage tank on a Safety Injection signal and automatically opening the containment sump suction valves during the recirculation phase of operation, t
APPLICA81LITY: MODES 1, 2, and 3*.
ACTION:
a.
With one ECCS subsystem inoperable, restore the inoperable subsystem to OPERA 8LE s'atus within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT STAN08Y within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , b.
In the event the ECCS is actuated and injects water into the Reactnr Coolant System, a Special Report shall be prepared and submitted to the Commission pursuant to Specification 6.9.2 within 90 days describ-ing the circurstances of the actuation and the total accumulated j
actuation cycles to date.
The current value of the usage factor for each affected Safety Injection nozzle shall be provided in this Special Report whenever its value exceeds 0.70.
a l
i I
l "The provisions of Specification 3.0.4 and 4.0.4 are not applicable for entry into Mode 3 for the centrifugal charging pumps and the safety injection pumps declared incperable pursuant to Specification 3.5.3 provided the centrifugal j
charging pumps and the safety injection pumps are restored to OPERABLE status within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or prisr to thi temperature of one or more of the RCS celd leg $
r exceeding 375'F, whichever comes first.
i w
I COMANCHE PEAK - UNIT 1 3/4 5-3
i.
III 0$fH &,bl69 PAsl 13 W 15 Ensste,caer con cowmu srsts e s gy, f
3/4.5.4 REFUELING WATER STORAGE TANK LIMITING CON 0! TION FOR OPERATION 3.5.4 The refueling water storage tank (RWST) shall be OPERABLE with:
Aminimumcontainedboratedwatervolumeof(479,900 gallons a.
((Later}5 of span),
Aboronconcentrationofbetween2000and[2200]ppeofbaron, b.
c.
f. minimum solution temperature of 40*F, and d.
A maximum solution temperature of 120*F.
APPLICA8!LITY: MODES 1, 2, 3, and 4.
ACTION:
With the RWST inoperable, restore the tank to OPERA 8LE status within I hour or be in at least HOT STAN08Y within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SRUTDOWN within the
/"
following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .
I SURVEILLANCE REQUIREMENTS i
4.5.4 The RWST shall be demonstrated OPERA 8LE:
l a.
At least once per 7 days by:
1)
Verifying the contained borated water volume in the tank, and 2)
Verifying the boron concentration of the water, b.
At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by verifying the RWST temperature when the outside cir temperature is isss than 40'F or greater than 120*F.
i CCMANCHE PEAK - UNIT 1 3/4 5-10
M M e,4,no CPSES Technical Specifications NRC Oraft 2 Markup Section 3/4.6 Chance 108 Justificatio1 For Chance 0268 Replace /Delute reference to Table 3.6-1 with the CPSES 0283 Technical Specification Improvement Program (TSIP).
0290 This change is based on the relocation of Table 3.6-1 0292 to the TSIP which is iustified under Soecification 0294 3.6.3.
5* mu n T W tt16o i
M 7 " "M 9
0296 u%t R21u, 0279 A note has been added to the Action S u tement to allow for the OPERABLE air lock door tnat is required to be closed by the A-tion Statement to be opened for short periods of time.
This note only applies to t'; outer air lock door and for a specified time of 15 minutes per individual entry.
This change would allow the inner door to be repaired and returned to an OPERABLE status with minor impact of safety and plant operations.
By following this Action there is presently no p-ovision to allow opening the air lock door that is required to be closed except to voluntarily enter 3.0.3 which is strictly not allowed.
This change is similar to that licensed at Vogtle, Shearon Harris and Seabrook.
0281 i
Action time requirement of one hour is not consistent GoMt%Mb lid MBBM0 with eight hours allowed by the containment sg un%mNT 9 temperatur ecification 3/4.6.1.5.
Both of these Tnc.s 4 A oF M parameters are u in the same accident analyses as a set cHmr.6 :cb401%)
of initial conditions, his additional time allowance would provide adequate ti perform all requ' red samples and fulfill all adminis ive requirements.
Without this change the entire proce o'J1d have to be started so early in the pressure band tha e operators would be unduly burdened which would distract from peforming their normal duties of maintaining the pla n
a stable and safe condition.
0282 This Surveillance Requirement has oeen revised 0485 to reflect the plant specific system design for average containment temperature.
The average containment temperature is determined by th'e volume weighted adjusted average tempe?ature of two temperatures of which one is from elevation 1000'6' or above and the other is from elevation 860' or above.
This adjusted average will be determined by correlating the preoperational test data te determine the correction factor to be used for the temperat.ures taken at the two different locations.
This will allow remote or local measurements of the containment temperature a; well as the flexibility of being able to use any location (s) above the specified elevations since one of tine two temperatures will always be determined using the correction factor for elevation 860. If, a higher elevation is used for this reading, the calculated
{,
average temperature would be higher than the actual average temperature since as the reading elevation is increased,the temperature increases.
III W 8 Af facwor,tito CPSE3 Technical Specifications nst 3 or y NRC Oraft 2 Markup Seccion 3/4.6 Change 10#
Justification For Chance 0285 i
C0 Statement.ind Action Statement for l
0286 the 1 h pressure relief valves have been changed to suntsmb 'B9 wx efMEO. all w f r e valves to be opened as necessary to relieve contain t pressure or for required surveillance g%gg ypgg testing.
This is b on the fact that the valve has (AAw Ibesuf been qualified to shut he time prescribed in the accident analysis.
This is omplished by physically limiting the valve opening to en that the valves meet isolation time requirements.
This p specific change is similar to that ifcensed at South Texa Shearon
(
Harris, vogtle and Seabrook, i
i 0288 See 108 0268 0290 See 10# 0268 0291 Added new Action e. to take exception to Specification 3.0.4.
If a containment isolation valve is inoperable, proper isolation of the affected penetration per Action Statements b or c. will maintain containment integrity 1
and allow operations to continue in Modes 1, 2, 3, and 4 with no impact on plant safety or (peration.
0292 See 10# 0268 0293 Delete the phrase "dsring COLO SHUTOOWN or REFUELING MODE" removes an unnecessarily resrictive requirement for the l
desired testing. Many of the valv?s will or could be t
tested for Phast "A",
Phase "B" isolation, or Containment Ventilation isolation during the quarterly Slave Relay l
Testing.
This change is similar to that Licensed at i
Olablo Canyon and Paterford.
t 0294 See 10# 0268 0296 See 10# 0268 l
0485 See 10# 0282 0575 The addition of a 3.0.4 exemption is consistent with the f
other post accident Technical Specifications.
The i
allowance to be able to change Modes while complying with the 30 day Action Requirement is the same as that allowed
~
by Technical Specifications 3.3.3.6, Accident Monitoring Instrumenta* ion, anc 3.6.4.2, Hydrogen Monitors, which both have 3.0.4 e>.enptions.
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h
[
Tw una M%'[g CPSES Technical Specifications NRC Oraft 2 Markup Section 3/4.6 i f Change 10j hstificationforChange Md1 The Action time requirement of one hour is not consistent with the eight hours allowed by the containment temperature Specification 3/4.6.1.5.
both of these parameters are used in the same accident analyses as a set of initial conditions.
The pressure band at CPSES is much narrower than most plants that have a large containment.
With this very narrow pressure band and the fact that the or.ly way to relieve the pressure will be through the containment pressure relief valves which only have an effectivo diameter of 3 inches, it is anticipated that this condition could become operationally constraining for no appdrent technical reason.
This additiocal time allowance would provide adequate time to perform all required samples and fulfill all administrative i
requirements. Without this chan e the entire process would have to be started so earl in the pressure band that the o)erators would be undu y burdened which would distract tiem from performing their nomal duties of maintaining the plant in a stable and safe conoition.
This justification supersedes the justification for 10 i
0281 in TV Electric Letter TXX-88512.
i 5117 By relocating Table 3.6-1, it also removes the means to identify exceptions to the Containment Integrity requirement.
Identifying the conditions as to when the 4
exceptioncanbeused(i.e.surveillancetestingornormal plant evolution) limits the time that these valves will be opened.
Normal plant evolutions is used here to mean evolutions that are performed using an approved procedure, j
This gives assurance that if these valves are opened thay will be shut in a reasonable period of time.
This justification supersedes the justification for 10 #0268 in TV Electric Letter TXX-88512.
5118 Changed the allowed accumulative opening time for the 18-inch containment pressure relief discharge isolation valves to be unlimited.
This is based on redesign of the system which limits the amount the valve will open and adds an orifice plate which changes the effective diameter of the valves to 3-inches from the combination of th",e modifications.
Therefore with the valves essentia /
being equivalent to a 3-inch valve and being capable of closure to meet the accident analysis assumptions any time restrictions do not apply.
This justification supersedes the justification for 10 #0285 and 10 #0286 in TV Electric Letter TXX-88512.
% x 60780 CPSES Technical Specifications i
"C""
N 4 I " u NRC Draft 2 Markup i
Section 3/4.6 (Cont'd) l Chang,q L0ft Justification for Chance 5120 By relocating Table 3.61, the notes attached to that table were also relocated.
In order to take exceptions to the LCO, the specific notes must stay in the Technical Spacifications. This note allows the operator to be able to follow a single technical specification for the givan system. The individual technical specifications include the same surveillance testing as 3,6.3.
5121 The containment purge and hydrogen purge valves are i
required to be maintained locked closed in Modes 1 through 4 by Specification 3.6.1.7.
Since Specification 3.6.3 is orly applicable in Modes 1 through 4 there is no reason to stroke test these valves for a maximum isolation time.
91X 96H SMGo AffAChPEN! 9 FAGE 5 Of 42
(
3/4.4 CONTAINMENT SYSTEMS 3/4.6.1 PRIMARY CONTAINMENT CONTAINMENT INTURITY LIMITING CONDITION FOR OPERATION 3.6.1.1 Primary CONTAINMENT INTEGRITY shall be maintained.
APPLICABILITY: MODES 1, 2, 3, and 4.
ACTION:
Without primary CONTAINMENT INTEGRITY, restore CONTAINMENT INTEGRITY within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months or be in at least HOT STANOBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hou.'s, hWs N t ses c de oiret h befec m-4 pe n h F e gw m svenA e te s t s e,
'T' M t (4 N tiu s m e, N.
ci, on u SURVE!LLANCE REQUIREMENTS MYU
^ '.
4. 6.1.1 Primary CONTAINMENT INTEGRITY shall be demonstrated:
a.
At least once per 31 days by verifying that all penetrations
not capable of being closed by OPERA 8LE containmont automatic isolation valves and required to be closed during accident conditions are closed by valves, blind flanges, or deactivated automatic valves secured in their positions, eme-" " ---""-'
'a " ' " ' * ' " "
+l O::t fic: tin 3. 5. t.1;+
dew : ! u = C+
= W, 0-f~~tttV e
I W
b.
By verifying that each containment air lock is in compliance,with I
the requirements of Specification 3.6.1.3; and Ib4 Sin c.
Af tar each closing of each penetration subject to Type 8 testing, except the contair.nent air locks, if opened following a Type A or 8 test, by leak rate testing the seal with gas at a pressure not less than P, 44.3 psig, and verifying that when the measured leakage rate forth$sesealsisaddedtotheleakageratesdeterminedpursuantto Specification 4.6.1.2d. for all other Type 8 and C penetrations, the combined leakage rate is less than 0.60 t..
t "Except valves, blind flanges, and deactivated automatic valv.s which are 1ccated inside the containment and are locked, sealed or otherwise secured in the closed position.
These penetrations shall be verified c1tsed during each COLD SHUTD0'nN except tt. '; such verification need not be performed more often than once per 92 days.
The blind flange on the fuel transfer canal need not be verified c.losed except after each drainage of the canal.
CCMANCHE PEAK - UNIT 1 3/4 6-1 t
!!1-MHF BM Be 3rrgegr 9 PAGEISCF42 CONTAINMENT SYSTEMS 7
CONTAINMENT VENTILATION SYSTEM LIMITING CONDITION FOR OPERATION
3. 6.1. 7 Each containment rd hydre;:n ventilation isclation valve shall be OPERA 8LE and:
a.
Each 44-inch and 12-inch containment and hydrogen purge supply and exhaust isolation valve shall be locked closed, and R4 F,l k.
AU W.
WUJ The 18-inch t.ontainment pressure relief discharge isolatic. vstve(s) d, t h-shalt b cPt u Bt.E t_ "aey- : :;;r. fw-w,-t 90 heur: durin
,M-E: A nte,, w
,d _g:, ::1:rd:r y:: L
~
q:c k ;2,g
+'-t
--... APPLICABILITY: MODES 1,2,3,andI.
_2;m.h.. br ymH a-n. a. ', _ j, A
.s,
. mehcc te m
^h C V ^ ^ ^ bi O l'U L E ACTION:
a.
With any 44-inch or 12-inch containment or hydrogen purge supply and/or exhaust isolation valve open or not locked closed, lock riose that valve or isolate the penetration (s) within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , otherwise be in at least HOT STAN08Y within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD 2b,,gs t),g SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .
,+e e.w, c U.se,
N.,
' * ^
- 3 L J 7 L M e, With the 18-inch containment pressure relief dlscharge isolation int *'*V8 %'b-valve (s%:;:r f+r F--- '"-^^ Nr: dur%M cel:rder year, close the rN*5 O N open Idinch valve (s) or isolate the penetratton(s) within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , MY #7D otherwise be in at least HOT STAND 8Y within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , and in COLD SHUTOOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .,
c.
With a contairment pressure relief discharpe isolation valve (s) having a measured leakage rate in excess of the Itaits of Specifications 4.6.1.7.3 or with the containment and hydrogen purge supply or exhaust isolation valve (s) having a measured lea, age rate in excess of the limit of Specification 4.6.1.7.+
restore the inoperable valve (s) to OPERABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 7,otherwise be in at least HOT ST/WOBY wit'inthenext6 hours,andinl' COLD 5HUTDOWNwithinthefollowing 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .
2.
SURVE!LLANCE REQUIREMENTS 4.6.1.7.1 Each 44-inch and 12-inch containment and hydrogen purge supply and exhaust isolation valve shall be verifica to be locked closed at least once per 31 days.
9 66 4-4-1-7-2 The wiethe-time-thet-eM-18-kch-pe*+sure relief 44th+ete J
M144 M-v+1e e: hove deer open duria; a celead=" ye="
e11 be det = 4aad 2+- h ".M lea:t Oste-ptr T d!y5.
l CCMANCHE PEAK - UNIT 1 3/4 6-9
fir WM speo Ali&CMENT 9 PAGE !? 0F 42 CONTAINMENT SYSTEMS 3/4.6.3 CONTAINMENT !$6LATION Vt.'.VES
(
QITINGCONDITIONFOROPERATION n e sito 3.6.3 Tha containwnt isolation valves etee"icd '- Tel: 3 F1 shall be OPERABLE.wf m-Mietien-times-as-:h wn-in Tele-M4,.
10 lj o2p AfPLICABILITY: MODES 1, 2, 3, and 4.
ACTION.
With one or more of the containmert isolation valve (s) ::::f f f;d in isli !)If 0290
-h6-+ ino9erable, a.aintain at ihast one isolation valve OPERA 8LE in each i
affectea penetration that is open and:
t.
Restore the inoperable valve (s) to OPERA 8LE status within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , or b.
Isolate each affected penetration within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months by use of at le'ast one deactivated automatic valve secured in the isolation position, or c.
Isolate each affected penetration within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months by use of at least one closed manual valve or blind flange, or
(~
d.
Be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .
s' 10 1:
0291 S,URVEILLANCE REQUIREMENTS i
4.6.3.1 The containment isolation valves :;::ified i-Tel: 3.P1 shall be demonstrated OPERA 8LE prior to returning the valve to service af ter maintcnance,gg i. 0292 t
l repair or replacement work is performed u. the valve or its associated actuator, control or power circuit by performance of a cycling test, and verification of isolation time.
The provistsns of Specification 3.0.4 are not applicable provided e.
that the affected penetration is isolate.f in accordance with ACTION b or c above, and provided that che associated system, if appitcable, is declared inoperable and the appropriate ACTION statements for that system are performed.
N ' CAUTION: The inoperable isolation valve (s) may be part of a systee(s).
Isolating the affected penetration (s) ma/ affect the use of the sys*,es(s).
Consider the technical specification recuirements on the affected system (s) and_act accordingly.
n a5iLe
[ Jt%e re remt of S pet. Cit u.e s 1.6 3 Ao not a.
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" W p(< n reyts% wt #,e a n a 7. i.,.3 us Q 37gg ec
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COMANCHE PEAK - UNIT 1 3/4 6-13
!!I Witt 64 r 60 AllACHMNT9 PAGE 24 0F 42 RE.00A"I.'" "'
TABLE 3.6.1 (Continued)
CONTAINMENT ISOLATION VALVES MAxfMuM ISOLATION TYPE FSAR TABLE TIME LEAK TEST yALVE NO.
REFENENCE NO.*
LINE OR SERVICE (Seconds) REQUIREMENTS 2.
Phase N " Isolation Valves 1HV-4708 117 C0 Return from RCP's 10 C
Motors 1HV 4701 117 CC Return from RCP's 10 C
Motors 1HV-4700 18 CC Supply to RCP's 10 C
Motors 1HV-4709 119 CC Return From RCP's 10 C
Thermal Barrier 1HV*4696 119 CC Return From RCP's 10 C
~
Thermal Barrier 3.
Containment Ventilattan Isola ion Valves no rip 1HV-5542 58 C.
drogen Purge Supply
)#'gA C 1HV 5543
$8 Hyd en Purge Supply WA'/A C INV-5563 58 Hydrog n Purge Supply g4/A C 1HV-5540 59 Hydrogen urge Exhaust gA/A C 1HV 5541 59 Hydrogen Pu e Exhaust W4/A C 1HV-5562 59 Hydrogen Purge Exhaust g4/A C 1HV 5536 109 Containment Purg Air gd4 C
Supply O
1HV-5537 109 Containeent Purge Ai g4/A C
l Supply 1HV-5538 110 Containment Purge Air ' g4/4 C
Exhaust 1HV 5539 110 Containment Purge Air
/A C
Exhaust 1HV-5548 122 Containment Pressure 3
Cs
(
Relief
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(
's COMANCHE PEAK + UNIT 1 3/4 6 20
tit 8titteE784 attat W I 9 MN0 1:1 OHO TABLE 3.6.1 (Continued)
TA8LE NOTATIONS (dentificationcodeforconttinsentpenetrationandassociatedisolation vatises in FSAR Trbles 6.2.4-1, 6.2.4-2, and 6.2.4-3.
Note These are closed systems which rett the requirements of NUREG-0800 Section 6.2.4, !!.6, paragraph o.
These valves are theretore not required to be leak tested.
N.ote 2:
T se valves inside containment are part of closed systems outside containment which are in service post accident at a pressure in exceh of containment design pressure and satisfy single failure criter' on.
These va*ves are therefore not required to be leak tested.
Note 3:
These are losed systems outside containment which are in service post accid t and have a water-filled loop seal on the containment side of the lves for a period greater than 30 days following the accident. Thdge valves are therefore leak rate tested with water at a pressure o P,.
I Note 4:
These ESF valves a normal.15 open and remain ope.n e ing post.
accident conditions.
Post $ccident they are con',inually pressurized l<L in excess of contal t pressure from an ESF :;surce which meets the single failure criterio. These valves are therefore not required i
to be leak tested.
(
Note 5:
An effective fluid seal on ese penetrations is provided by the i
A suction sources to the residu 1 heet removal pumps during and fol-I lowing an accident.
In additi
, these containment isolation valves l
are non automatic, are not requi d to operate postaccident and are located inside containment.
Thes valves are therefore not required i
j to be leak tested, t
Note 6:
All four MSIV bypass valves are locked losed in Mode 1.
Duriirg l
Modes 2, 3, and 4 or.e MS!V bypass valve.may be opened provided l
theotherthreeMSIVbypassvalvesarelo%edclosedandtheir associated MSIVs are closed.
l 1
l Note 7:
TheseareparallelESFvalvesthatarenohuill closed, but are designed to open during post-accident condition Failure of one t
valve to open will not prevent systes pressurizat n on both sides of both valves in excess of containment pressure, hose valves are therefore not required to be leak tested.
i I
Note 8:
These valves located outside containment are normally c sed and seeapressureinexcessofcontainmentpressureinpost-\\ccident l
conditions.
A valve sten leakage check will be performed to a i
quarterly basis to assure no significant stee leakage would ccur
~
in post accident conditiona art Note 9:
These valves require steam to be tested 4re are thus not require to lt 8
/
be tested untti the plant is in MODE 3.
\\
i CO
NCHE PEAK - UNIT 1 3/4 6 31 I
II! Wit 88780 AllACHMENT9 PAGE C C' 42
~
~
plrl CONTAINMENT SYSTEMS BASES I4 1: 048$
3/4.6.1.5 AIR TEMPERATURE The limitations on containment average air temperature ensure that the overall containment average air temperature does not exceed the initial tem-perature condition assumed in the safety analysis for a LOCA or steam line break accident. -44e**vrements-sh*1' 50 :d:-et :11 li:ted Ie;;thn:, wh:thec ID 1: 0445 by-fheG er-;ertetde-bstrumentsT 1.rier te deMicdrJ g k' tie e,ecege e're t-t retur,, % c as * <
y t e-t er =% e t LD o
at \\ c. o. O
?. ok ut m e a.sa r e m e rb m eAe. a.4 We gi te d.d ed, a,ver6p g an lcc t: sng 3/4 6.1.6 CONTAINMENT STRUCTURAL INTEGRITY D* *
l * * 't'd"N " \\"%
C'M *h d e c o m.e. b c temprs 6e roaaw rt me nt-acen. m.
This limitation ensures that the structural integrity of the containment will be maintained comparable to the original design standards for the life of the facility.
Structural integrity is will withstand the maximum pressure of,reqw red to e'isure that the containment L48.3, psig in the event of a LOCA.
A visual inspection in conjuncti:n with the Type A leakage tests is sufficient to demonstrate this capability.
~
3/4.6.1.7 CONTAINMENT VENTILATION SYSTEM
~ (
The 46-inch and 12-in:h containment and hydrogen purge supply and exhaust
\\.
isolation valves arn required to be locked closed during plant operations i
since these valves have not been demonstrated capable of closing during a LOCA er steam line break accident.
Mai'itaining these valves locked closed during plant operation ensures that excessive c..lantities of radioactive materials will not be released via the Containment Ventilation System.
To provide assurance that these containment clives cannot be inadvertently opened, the valves are locked closed in accoidance with Standard Review Plan 6.2.4 which includes mechanical devices to seal or lock the valve closed, or orevents power from bhgelied to the valve oDorator.
. 7-i
<,_t W th an a%%e Atamder *I 3 Mu) n 45'HB
" '7 TheuseoftheContainmentVent114tik$ystemcuringoperationsis G ese restiicted to the 18-inch pressure relief discharge isolation valves *since h ve n t. o 444e-N '*-t-e :nd 12 JM5 velvet,-the 18 iac" valves are capable of.
. closing during a LOCA or steam line break accident.
Therefore, the GIE-OMue4 AcrA M M e"7 dose quicalir.a of 10 CFR 100 woulo not ce exceeded in the event of an
%N81 accident during containment,v,eriTirg operation.
Ope <4t44 eith one pet-of these-valv:: 0;;r =64}-b: l'=ited t: 90 ' ;r: dur'ng : : hnd:r y::r
'50 total-t4me-the-<entainment pur;; 'v:nt) y:t:: H hthn c:h:: ny b: caen C ri ; WCES 1, 2, 3, 2nd ' i
: !:nd:r y::r b : fun:the Of anti +4peted n
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need ed creratt + sapu. ten e.
Oaly safeterelated eascas; e4,_ caateh=ent pressure-conteel-ee-two-redwet4e*-+4-*irt:rn; redice:t4vity te f act++ tete per44anet-access-fon-erve!' un:: :nd ::4*teeeeee-eetivities, iney Le neG i.e support-the--44dt t hn:1-tim r:;-e:ts. Caly 'efety-re hted 7:00:n;dne h b;
-esed te je:44fy th: 0;;r k; cf-the:: huht4:n v;h:: :;r'ngy0Cyl.2d, en 4.M :ny-celeedae-yene-regaeele : Of th; OII:100 k -*evet.\\.
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CPSES Technical Specifications NRC Oraft 2 Markup
~
Section 3/4.7 1
Change 10#
Justification For Chance (cont.)
The downstream HEPA filter bank was added to the design of cleanup trains many year ago, p;imarily to trap any carbon fines that may be released from the adsorber beds.
Some designers felt this would also be a good back-up to the upstream HEPA bank in the event of total failure of that bank.
This has since been accepted as not a credible scenario.
The installation of two filter banks in one housing was not intended to satisfy any single-failer.
proof criteria.
l A high-grade 95% roughing filter would be sufficient to trap carbon fines.
The minimum carbon particulate size to be installed in a bed is governed by qualification testing prior to installation.
Change-out of the HEPA filters presently installed, to roughing filters, is not being advocated due to seismic and pressure dro) considerations.
The downstream HEPA filters tend to 1 ave an exceptionally long service Itfe l
and should only degrade from accidental damage during maintenance activities or loading due to a dirty filter housing.
This change is made to make the Technical Specifications consistent with the FSAR.
0311 Added footnote to clarify which revision of ANSI N 510 0315 referenced by Regulatory Guide 1.52 will be used.
0323 This is consistent with the commitment made in the 0324 FSAR, Section IA/B.
0328 su ne.u u b M nx.-ea7so 0329-M eracAmec io 0491 PRn soA or 54 r
0492 cmu Ib 4 512.2.
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0314 See IDF 0310 0315 See 107 0311 b
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CPSES Technical, Specifications NRC Oraft 2 Markup Section 3/4.7 Change 10#
Justification for Change L5122 Delete the reference (*) to ANSI N510 since this standard does not deal with charcoal analysis.
The reference in Regulatory Guide 1.52, Revision 2, March 1978 references the correct standard to be used for this analysis.
This justification suparsedes the justification for 10 #0311, t
10 #0328, and 10 a0329 in TV Electric Letter TXX-88512.-
5123 The CPSES design has changed to where the Plant Vent High Radiation signal no longer is taken credit for in the scenario requiring control room emergency recirculation.
This signal is being replaced with redundant Intake vent-High Radiation signals.-
i i
f i
i 4
Ill*WM $6180 Af fMEMNi lo PAGE 21 # 54 s
PLANT SYSTEMS STEAM GENERATOR ATMOSPHERIC RELIEF VALVES LIMITING CONDITION FOR OPERATION 3.7.1.7 At least four atmospheric relief valves and associated remote manual controls shall be OPERABLE.
APPLICABILITY:
MODES 1, 2 and 3.
ACTION:
a)
With one less than the required atmospheric eteenr l (b.
relief valves OPERABLI, restore the required atmospheric st ca-relief valves to OPERABLE status l0 within 7 days; or be in at least HOT STANDBY vithin the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and place the required RCS/RHR loops in operation for decay heat removal, b)
With two less than the required atmospheric relief valves OPERABL2, restore at least three atmospheric relief valves to OPERABIZ status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT STANDBY vithin the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in
!!OT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and place the required RCS/RHR loops in operation for decay heat removal.
SURVEILLANCE REQUIREMENTS 4.7.1.7 Each atmospheric relief valve and associated manual controls shall be demonstrated OPERABLE by a)
At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by verifying that the air -nitr:g:n accumulator tank is at pressure greater than or equal to (80) psig, b)
Testing pursuant to Specification 4.0.5.
3/4 7-S c
III M Wt 88160 I
AffACH5NT10 l
FAGE 26 of $&
I PLANT SYSTEMS 3/4.7.6 FLOOD PROTECTION LIMITING CON 0! TION FOR OPERATION 3.7.6 Flood protection shall be provided for all safety-related systems, components, and structures when the water level of the Squaw Creek Reservoir (SCR) exceeds 777.5 Mean Sea Level USGS datum.
l (O' f st.t d APPLICABILITY: At all times.
ACTION:
With the water level of SCR above elevation 777.5 Mean Sea Level USGS datum, l(0-initiate and complete within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , the flood protection measures verifying that any equipment which is to be opened or is opened for maintenance is isolated from the SCR by isolation valves, or stop gates, or is at an elevation above 790 feet.
SURVEILLANCE REQUIREMENTS 4.7.6 The water level of SCR shall be determined to be within the limits by:
a.
Measurement at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months when the water level is below
(
elevation 776 Mean Sea Level, USGS datum, one.
hb.
b fett b.
Measurement at least once per 2 hout s when the water level is equal to or above elevation 776 Mean Sea Level. USGS datum a.edL.
3 t b' t
- tss L l
c.
With the water level of SCR above 77 an Sea Level. USGS datum, verify flood protection measures are in effect by verifying once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months that flow paths from the SCR which are open for' maintenance are isolated from the SCR by isolation valves, or stop gates, or are at an elevation above 793 feet.
j e
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COMANCHE PEAK - UNIT 1 3/4 7-13 d
a
Vn-tntt 86"!BC AttACMEN710 PAGE880F54 PLANT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) b.
At least once per 18 months or G) af ter any structural maintenance on the HEPA filter or charcoal adsorber housings; n m failewina_
pefnting rffreroe-chemica1-re4eese-fn eny aatHetfon-rene-c;=#
u. ting with the system-by:
1)
Verifyiy that the filtration unit satisfies the in place pene-tration and bypass leakage testing acceptance criteria of less than 0.05% by using the test procedure guidance in Regulatory
(, vg em M P9_litlELMa, C.5.c4 )and C.S d of Regulatory Guide 1.52,
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Revisions 2. March 1978 and the emergency filtration unit flow rate is 8000 cfm 2 10%, and the emergency pressurization unit flow rate is 800 cfm 210%;
M h MM 2)
Verifying, within 31 days after removal, that a laboratory analysis of a representative carbon sample obtained in accor-dance with RegulatoryJosition C.6.b of Regulatory Guide 1.52,g y, p_
'tevision 2 March 1978%, meets the laboratory testing criteria -
of Regulatory PosLtion C.6.a of Regulatory Guide 1.52. Revi-sion 2, March 1978%, for a methyl iodide penetration of less than 0.2%; and M h M09 1)
Verifying h mergency filtration unit flow rate b 8000 ctm 110% and in mergency pressurization unit flow rate 4f 800 cfm 2 10% d W g :y:t:: ;;;retirm when tested in accordance with ANSI N510-19M. nuo.
t w.n p.st L
er chem;e,J Af ter every 720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months of charcoal adsorber operatio or fo\\1h'*1 ST%A f *% h*, N c c.
,b
rYfying,
& n.c within 31 days after removal, that a laboratory analysis of a repre-sentative carbon sample obtained in accordance with Regulatory m 0 0-Position C.6.b of Regulatory Guide 1.52, Revision 2. March 19't meets the laboratory testing criteria of Regula ory Position C.6.a of Regulatory Guide 1.52, Revision 2, March 1978, for a methyl IWlu todido penetration of less thaa 0.2%;
~
d.
At least once per 18 months by:
1)
Verifying that the pressure drop across the combQed]HEPA filters und charcoal adsorber banks is less thant/.7 inches Water Gauge while operating the emergency filtration unit at a flow rate of 8000 cfm Gauge while onrating ! 10%, and is less than[9.25J inches Water the emergency pressurization unit at a flow rate of 800 cfm 2 10%;
g,, y n 2)
Verifying that on a Safety Injection, Loss of-Offsite Power,cr-Intake Vent-High Radiation, er "hnt Vent 44 h-44444tt+a-test-t
-+4aeh the train automatically switches into the emergency recirculation moce of operation with flow through the HEPA filters and charcoal adsorber banks; s
3)
Verifying that the emergency pressurization unit maintains the control room at a positive pressure of greater than or equal
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b AMT 4 S IC-1%O A d be ded. m p\\ e oh S 45I N5 Q 1 1 7.L Is zr c )t5
N-tM CCMANCHE PEAK - UNIT 1 3/4 7-15
III 96M 8818e AffAtxM(WT10 PAGE 30 Of 54 l
PLANT SYSTEMS 3/4.7.8 PRIMARY PLANT VENf!LATION SYSTEM - ESF FILTRATION UNITS LIMITING CONDITION FOR OPERATION 3.7.8 Two independent ESF Filtration Units shall be OPERABLE.
APPLICABILITY: MODES 1, 2, 3, and 4.
ACTION:
With one ESF Filtration Unit inoperable, restore the inoperable ESF Filtration Unit to CPERABLE status within 7 days or be in st least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .
SURVEILLANCE REQUIREMENTS 4.7.8 Each ESF Filtration Unit shall be demonstrated OPERABLE:
At least once per 31 days on a STAGGERED TEST BASIS by initiating, a.
from the control room, flow through the HEPA filters and charcoal
(
adsorbers and verifying that each ESF Filtration Unit operates for at least 10 continuous hours with the heaters cperating; b.
At least once per 18 months or 44) after any structural maintenance on the HEFA filter or charcoal adsorber housingst-er (2)-f 11M n; pefetiem-f fr;, or ch;dcal-Miease-insey-venti 4*t4en-teet-een -
municate; eith the synes by:
!D I: 0321 1)
Verifying that each ESF Filtration Unit satisfies the in place penetration and bypass leakage testing :ccept:nce criteria of l
less than 1.0% by using the test procedure guidance in Regula-
.h e r WP_.1_ Revision 2, March 19 81 and verifying the flow rate is 15,00 tory Positions C.5.a C.5 4 and C.5.d of Regulatory Guide 1.52, +
(o y
M
10% per ESF Filtration Unit when tested in accordance with ANSI N510-19M; and l
sqsc 10 1: 0324 to I. 0322 '
2)
Verify ng, within 31 days af ter removal, t iat a laboratory analys s of a representative carbon sample obtained in accor-dance with Regulatoryh meets the 1:,,aratory testing crite Revision 2, March 1978 l
of Regulatory Pos ion C.6.a of Regulatory Guide 1.52, Revi-sion 2 March 197 for a methyl iodide penetration of less than 1.0%.
-V - A>4 ST N Sc -l%c shdl be ul. W luc. d A>)sI 4 5io.117.5.
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10 : 0323 COMANCHE PEAK - UNIT 1 3/4 7-17
TWN5tf $8 rto..__
i ATTACMENT 10 PAGE 31 CF 54
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or ko! Icd.n3 p,nlin3 d irt 'a.o r c.htmied b
3l N.\\ u% t m co mm e a dy\\ d ie rtca.ny u;v s.nt;l k.oa,EcNe yJ i
I PLANT SYSTEMS g g y
p a.n b V v. et '.
s p te.rn 10 8: 0327 4
s SURVEILLANCE REQUIREMENTS (Continued)
Afterevery720hoursofcharcoaladsorberoperatio8,byverifying, c.
within 31 days after removal, that a laboratory analysis of a repre-
., g, ser,tativt carbon sample obtained in accordance with Regulatory s Poiltio.1 C.6.b of Regulatory Guide 1.52, Revision 2 March ifl78N. U 8-M meets the laboratory testing criteria of Regula ory Pasition C.6.a of Regulatory Guide 1.52, Revision 2, March 1978, for a nethyl d odido penetration of less than 1.0%;
At least once per 18 months by:
1)
Verifying that the pressure drop a:ross the comb ed H, EPA filters and charcoal adsorber banks is less than 8.25jinchesitataflowl Water Gauge while operating each ESF Filtration rate of 15,000 cfm + 10%,
2)
Verifying that each ESF Filtration Unit starts on a Safety Injection test signal, and 3)
Verifying that the heaters dinsipate 1"' + 5 kW when tested in i
accordance with ANSI N510-M75.1980 i
y g
e.
Af ter each complete or partial replacement of -*'HEPA filter bank, by verifying that the associated ESF Filtration Unit satisfies the 2
in place penetration and bypass leakage testing acceptance criteria g
of less than 1.0% in accordance with ANSI N510-19&for a 00P test
,qe aerosol while, operating the associatea ESF Filtration Unit at a i
flow rate of 15,000 cfm i 10%; and f.
After each complete or partial replacement of a cnarcosi s.orber bank, by verifying that the associated ESF Filtration Unit,atis-fies the in place penetration and bypass leakage testing acceptance l
criteria of less than 1.0% in accordance witt "4SI N510-197hfor a
'I O halogenated hydrocarbon refrigerant test gas wr.ile operating the anociated ESF filtration unit-:y:t : at a flow rate of 15,000 cfm l
t 10%.
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CCMANCHE PEAK - UNIT 1 3/4 7-18
l IIIteWe naso AffACMENT 10 PAGE 50 Of 54 l
PLANT SYSTEMS BASES ULTIMATE HEAT SINK (Continued) l The limitations on minimum water level is ba.9d on providing a 30-day cooling water supply to safety-related equipment without exceeding its design basis temperature and is consistent with the recommendgtions of Regulatory Guide 1.27, "Ultimate Heat Sink for Nuclear Plants,", dv. 2 (January 1976).
The limitation on maximum temperature is based on the maximut allowable :ompo-nent temperatures in the Service Water and Component Cooling Water Systrms, and the requirements for cooldown. The 11mitation or, average sediment depth is based on the possible excessive sediment buildup in the service water intake channel.
3/4.7.6 FLOOD PROTECTION The limitation of flood protection ensures that facility protective r
actions w ll be taken in the event of flood conditions. The only credible i
Sleod feed condid on that endangers safety related equipment is from water entry (D.
l it to the turbine building via the circulating water system from Squaw Creek Reservoir and then only if the level is above 778 feet Mean Sea Level. This corresponds to the elevation at which water could enter the electrical and control building endangering the safety chilled water system.
The surveillarce requirements are designed to impleoent level monitoring of Squaw Creek Reservoir i
should it reach an abnormally high level above 776 fest.
The Limiting Condition for Operation is designed to implement flood protection, by ensuring no open f16w path via the Circula"-
'fater System exists, prior to reaching the nostulated flood level.
3/4.7.7 CONTROL ROOM HVAC SYSTS.M The OPERA 8ILITY of 'he f.ontrol Room HVAC System ensures that: (1) the 1
control room aseient air temperature does not exceed the allowable temperature per 3/4 7.11 for continuous-duty rating for the equipment and instrumentation l
i 2
cooled by this system, and (2) the control room will remain habitable for opera-tions personnel during and following all credible accident conditions. Opera-tion of the s', stem with the heaters operating to maintain low humidity using automatic control for at least 10 continuous icurs in a 31-day period is j
sufficient to reduce the buildup of moisture on the adsorbers and HEPA filters.
7be OPERA 81LITY of this system in conjunction with control room design provisions u based on limiting the radiation exposure to personnel occupying the control 1
l room to 5 rems or less whole body, or its equivalent.
This limitation is consistent with the, requirements of General Design Criterion 19 of 10 CFR 50 Appendix A.
ANSI N510- W M will be used as a procedural guide for surveillance i
{
testing.
1960 a b 001
(
CCMANCHE PEAK - UNIT 1 B 3/4 7-4
TII 80 tit &B780 AliACMENT10 PAGt 53 0F 56 PLANT SYSTEMS
\\
BASES i
3/4.7.10 SEALE0 SOURCE CONTAMINATION The limitations on removable contamination for tources requiring leak testing, including alpha emitters, is based on 10 CFR 70 39(a)(3) limits for plutonium.
This limitation will ensure that leakage from Byproduct. Source, and Special Nuclear Material sources will not exceed allowable intake values.
Sealed sources are classified into three groups according to their use, i
with Surveillance Requirements commensurate with the probability of damage to a source in that group.
Those sources which are frequently handled 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 or boron measuring devices) are considered to be stored and need not be tested unless they are removed from the shielded mechanism.
3/4.7.11 AREA TEMPERATURE MONITORING The limitations on nominal area temperatures ensure that safety-reitted equipment will not be subjected to temperatures that would impact their environ-mental qualification temperatures.
Exposure to temperatures in excess of the saximum temperature for normal conditions for extended periods of time could reduce the qualified life or design life of that equipment.
Exposure to tem-peratures in excess of the maximum abnormal temperature could degrade the operability of that equipment.
2.
3/4.7.1( UPS HVAC SYSTEM O-The OPERABILITY of the UPS HVAC System ensures that the l
uninterruptible power supply and distribution rooms ambient 10 h OD0 air temperatures do not exceed tne allowable temperature per 3/4.7 7 for continuous-duty rating for the equipment and instrumentahion cooled by this system.
' LL (V
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CCHANCE PEAK - UNIT 1 B 3/4 7-7
TII tetti 8643
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affacMDIT10 FAGE 54 0F 54 PLANT SYSTEMS BASES 3/4.7.1 SAFETY CHILLED WATER SYSTF4 b.
fhe OPERABILITY of the Safety Chilled Water System ensures that sufficient cooling capacity is available for continued operation of equipment dw4*g
.c c i d=
nditi

n;.
The redundant cooling capacity of this system, assuming a single failure, in consistent with the assumptions used in the safety analyses.
(AneY S The Safety Chilled Water System (SCWS), in conjunction with respective l
emergency fan coi P, is required in accordance with Specification Definition
!H 0341 1 20 (OPERABILITY) to provide heat removal in maintaining the various Engineered Safety Features (ESFs) room space de ipn temperatures below th,e,
associated equipment qualification limits for the range of Design Basis Accident conditions. Action 1equirements are provided to ensure OPERABILITY of the vital bus inverters and emergency battery chargers, by verifying within one hour that the normal HVAC system is providing space cooling to the vital power distribetion rooms.
The Action Requirement is provided to establish
]
within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months OPERABILITY of the Emergency Core Cooling Systems (ECCS) which do not depend on the inoperable SCWS.
The 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months period provides a reasonable time in which tr. establish OPERABILITY of this complement of key safety i
systems.
This requirement ensures that a functional train of ECCS equipment is available to put the plant in a safe, stable condition for the most probably abnormal operational occurrences.
An Action Requirement of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months l
is provided to establish OPERABILITY of the remaining required safety systems which do not depend on the inoperable SCWS. A seven day Action Requirement is for a single SCWS out of service, based on the high reliability of offsite power.
The ters "verify" is used in this context to determine if certain components are out-of-service for maintenance or other reasons.
It does not mean to perform the Surveillance Requirements needed to demonstrate OPERABIt!TY of the component.
3 l
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TIUN$ttegrgo
'""I U
W'"N CPSES Technical Specifications NRC Oraft 2 Markup Section 3/4.8 Change iD#
Justification For Change 0390 Delete reference to Table 3.8-1.
This chan ontherelocationofTable3.8-1totheTS!heisbased 0392 which is 0393 iustified below.
0395 hTcription of the function test is revised to reflect o
soM.Ssf.hw M m EA'l60 the previ a reed upon surveillance from CPSES Final j
%! manet"tt Draft Technica cifications.
The specific values for 4
7 %t 9B c5 M-testing acceptance c ia is more appropriately cunast %b d Si %
described in plant survel rocedures where the l
specific manufacturer's recommen for this acceptance criteria can be maintained and kept up te.
This change is similar to that Licensed at Wolf Cr=
Callaway, Vogtle, San Onofre, Byron, Catawba and Mc 0396 This TS is being relocated to the CPSES Technical Specification Improvement Program since precedence for complete removal of the TS has not yet been set.
TU i
Electric believes the inclusion of this TS is unnecessary l
and the requirements would..be.more appropriately addressed.'
in the CPSES Technical Specification improvement Program, however, only the table is being proposed for relocation at this time.
i Relocation of this specification is consistent with the guidance provided in the NRC's Interim Policy Statement (52FP3788), Febru ey 6, 1987, and the recommendations of
[
he Westinghouse Owners Group MEr.ITS Program.
l Priority is given to the relocation of this table since j
the detailed nature of Qis tchle would require fonnal license amendment = for such th ngs as a nomenclature i
change, a change in v endor supg. lier, a design improvement by a vendor, or plant modifications. Changes or modifications to equipment are already under the control of 10CFRSO.59.
The information currently in this table is more appropriately maintained in a document subject to TV Electric administrative control and 10CFR50.59 review i
under the CPSES Technical Specification improvement i
Program.
i The specific information supplied in Table 3.8 1 has not I
completed the validation process and is expected to change.
As soon as this information is available it will be transmitted to the NRC under a separate letter, ihis change is similar to that Licensed at Seabrook, Vogtle, Shearon Harris and South Texas.
fr H e6tt s e m atraostui11 CPSES Technical Sp;cifications PAGE 8 # 52 NRC Oraft 2 Markup Section 3/4.8 Change 10#
Justification For Change s
0927 Changed Hot "Shutdown" to Hot "Standby" in Action Statements a.,b., and c. to be consistent with the rest of the Action Statements and to ensure that their is no confusion as to how long the operator has to gat to Hot Standby between Action Statements.
0928 Citanged frequency from 31 days to 92 days for check / removal of wati.r in the fuel oil storage tanks.
% m u ub hf m aiso.
N hk mamut u The applicable requirement from the Standard Technical MG d W 52-Specifications for CPSES is 92 days because the N* *D#SM groundwater level is belew the level of the fuel oil storage tanks. N his significantly reduces the probability ofwaterincursic{(ttfactory.
into the storage tanks and 92 day verification is sa Performance of this surveillance requires hysical access to the fuel oil storage tank by removal f the missile shield covering the manways.
This evolution requires significant effort and 4
lowers the tank integrity wh' e the shield is removed.
This is not warranted due to t low probability of water entry to the tank. Note that no al fuel sampling from the tank is accomplished by pumpin fuel oil with the fuel oil transfer pump and doesn't requ re qhysical tank access.
In addition, if water was to bb Idup sufficiently in the fuel oil storage tank sludge pit t be pumped to the day fuel tank, the water wculd be evide in the 31 day check required by Specification 4.8.1.1.2b gr the fuel sample required by Specification 4.8.1.1.2e.
N 1
0929 Change 440 rpm to 441 rpm (typo).
Add (58.8 Hz) after the 441 rpm requirement to allow use of the frequency indication to perform surveillance.
4 j
120 (freq)
= RPM 120 (f) = 441
(# of poles) 16 f = 58.8 Hz
[
l 0930 ttwult_d wording in Action Statements a.1 and a.2.
The smcuttb M T(vatuo. changeNaa to allow removal of fuses to act as the w a w amtor o redundancy for tion isolation of an inoperable Q 4 A or 52 protective device.
t Qally, the breaker wording has m.w; zb w oGo been removed since a breaker is lly a subset of the protective devices and the word "bac s been replaced with "associated" to prevent confusion as to the j
table refers to as backup.
1
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Ttx.- Baloo mu.3mx il CPSES Technical Specifications 7er.4 4A er 52.
NRC Draft 2 Markup Section 3/4.8 Change 10#
Justification for Change 0930 Changed wording in Action Statements a.1 and a.2.
The change is made to allow removal of fuses to act as the redundancy for penetration isolation in the same manner as a breaker is tripped on an inoperable protective device.
Additionally,the "breaker" wording has been removed since a breaker is really a subset of the protective devices since CPSES uses both breakers and fuses for protection.
The word "backup" has been replaced with "associated" to prevent confusion as to what the table refers to as backup overcurrent protective devices. This justification supersedes the justification for 10 #0930 in TV Electric Letter TXX-88512.
5125 The Diesel fuel Oil (OF0) sampling and removal of water has been split up into two separate requirements.
The sampling will be performed on a monthly frequency by taking a sample off of the transfer pump discharge which means the sample will be representative of the fuel that the diesel will be using.
The CPSES design has the suction of the fuel transfer pumps six inches off the bottom in addition to the sludge pit to collect water before the water would interfere with diesel performance.
The real concern is not necessarily that water may get into the fuel but that micro-bioloolcal granth will occur at the OF0/ water interface and get sucked into the system which would clog filters / strainers and restrict flow of DF0 to the diesel.
This is not a concern at CPSES since microbiocide is added to the DF0 to prevent micro-biological growth.
By having 600 gallons of unusable volume it is anticipated it would take a significant failure to accumulate that amount of water and water would be identified during the monthly sample. All water will be removed every 18 months.
This change is also necessary based on manpower required to perform the sample and the potential for fuel contamination is significantly increased based on past experience.
In order to get all the water out of the fuel oil storage tank, the manway must be removed which leaves a 3 foot diameter opening in the tank.
This increases the likelihood of fuel oil contamination.
This justification supersedes the justi-fication for 10 #0928 in TV Electric Letter TXX-88512,
'nct-887 63 AttM4mcer it CPSES Technical Sp;cifications TMG in or st NRC Oraft 2 Markup Section 3/4.8 (Cont'd)
Change 10#
Justification for Change 5126 The present surveillance requirements for molded case circuit breakers does not adequately address the various types of breakers at CPSES.
If forced to use the STS, CPSES would be unable to comply based on the +20% testing criteria which in reality should only pertain ~to the adjustable setpoint breakers.
The non-adjustabla setpoint breakers should be tested to the criteria set by NEMA AB 2-1984 which is where the inserted sentence came from and therefore should be incorporated into the Technical Specifications.
This pro)lem was recognized at Seabrook and South Texas which in both cases the STS were changed.
This justification supersedes the justification for 10 #0395 in TU Electric letter TXX-88512.
HI B e tt 8 0 4 0 41TACFMNT 11 9 %E 10 of $2 i
3/4.8 ELECTRICAL POWER SYSTEMS 3/4.8.1 A.C. SOURCES af ""
OPERATING LIMITING CONDITION FOR OPERATION
3. 8.1.1 As a minimum, the following A.C. electrical power sources shall be OPERABLE:
a.
Two physically independent circuits between the offsite transmission networ*. and the onsite Class 1E Distribution System, and b.
Two separate and independent diesel generators, each with:
1)
A separate day fuel tank containing a minimum volusa of 1440 gallons of fuel, 2)
A separate Fuel Storage System containing i minimum volume, of
[88,175] gallons of fuel, and
( d.
3)
A separate fuel transfer pump.
APPLIC.91LITY: MODES 1, 2, 3, and 4.
ACJIC.
C
~.
a.
With one offsite circuit of the above-required A.C. electrical power sources inoperable, demonstrate the OPERA 81LITY of the remaining A.C.
sources by performing Surveillance Requirement 4.8.1.1.1.a within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and at least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months thereafter.
If either diesel generator has not been successfully tested within the past 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , demonstrate its OPERABILITY by performing Surveillance Requirements i
4.8.1.1.2.a.4 and 4.8.1.1.2.a.5 for each such diesel generator, ni on; renu separately, within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .g Restore the off ite; circuit to OPERABLE i
status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT SHUTOC"" within the next 6 -M hours and in COLD SHUTOOWN within the following 44-hours.
So b.
With either diesel generator inoperable, demonstrate the OPERABILITY of the above rwquired A.C. offsite sources by performing Surveillance Requirement 4.8.1.1.1.a within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and at least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months thereafter.
If the diesel generator became inoperable due to any cause other thar preplanned preventive maintenance or tes'.ing, demon-strate the OPERABILITY of the remaining.0PERABLE diesel ge.
wor by performing Surveillance Requirements 4.8.1.1.2.a.4 and 4.f. 1.2.a.5 wi').in 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months unless the diesel is already operating and Icaded #
<a: tore t',c inoperable diesel generator to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or ba in at least HOT !M99# within the next 46 hours5.324074e-4 days 0.0128 hours 7.60582e-5 weeks 1.7503e-5 months and in COLD SHUTOWN within the fo11owingf44. hours.
4 vau-30 IM M I tl --*
n 4: 0347
This test is required to be completed regardless of when the inoperable diesel go.ierator is restored to OPERABILITY.
i
0uring performance of surveillance activities as a requirement for ACTION stateunts, the air-roll test shall not be performed.
1 CCMANCHE PEAK - UNIT 1 3/4 8-1 l
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PA E 15 0F $2 ELECTRIC'ALPOUERSYSTEMS w
$URVEILLANCE REQUIREMENTS (Continued)
At least once per 31 days and af ter each operation of the diesel b.
whero the ceriod of operation was greater than or equal to I hour by checking for and removing accumulated water from the day fuel tank /;
9k it m e r, t M b i
3 1;; 0020 c.
At least t.nce per K ty: by cMch4 -fee-ane removing accumulated 9
water from the fuel oil storage tanks; g
n # sus I
d.
Oy sampling new fuel oil in accordance with ASTM-04057-1981 prior o addition to vtorage tar.ks and:
1)
By verif>ing in accordance with the tests specified in ASTM-0975-1981 prior to addition to the storage tanks that the sample hat:
a)
An API Gravity of within 0.3 degrees at 60'/, or a speci-fic gravity of within 0.0016 at 60/60*F when compared to the supplier's certificate, or an absolute specific gra-vity at 60/60'F of greater than or equal to 0.83 but less than or equal to 0.89, or an API gravity of greater than or equal to (26) degrees,but.less than or equal to (38]
degrees; b)
A kinematic viscosity at 40*F of greater than or equal to 1.9 centistokes, but less than or equal to 4.1 centi: nes C
(alternatively, Saybolt viscosity, SUS at 100'F of greate?
than or equal to 32.6, but less than or equal to 40.1), if gravity was not determined by comparison with the sup-4 i
plier's certification; i
c)
A flash point equal to or greater than 125'F; d)
A clear and bright appearance with proper color when tested in accordance with ASTM-04176-1982; 2)
By verifying within 30 days of obtaining the sample that the 1
other properties specified in Table 1 of ASTM-0975-1981 are met when tested in accordance with ASTM-0975-1981 except that the analysis for sulfur may be performed in accordance with ASTM-01552-1979 or ASTM 02622-1982.
I b e s a. 5 e.
At least once every 31 days by obtaining a sample of fuel oil in accordance with ASTM-02276-1978, and verifying that total particu-late contamination i. less than 10 mg/litet when checked in accor-dance with ASTM-022,'*1978, 6
Method A; w cr.e u,
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f.
At least once per 184 days by:
1)
Verifying the diesel starts from ambient conditions and the J
generetor voltage and frequency are 6900 + [690] volts and 60 +
1.2 Hz within 10 seconds after the start signal.
The diesel -
~4 generator shall be started for this test by using one of the signals listed in Surveillance Requirement 4.8.1.1.2a.4).
'i 2)
Verifying the generator is synchronized, loaded to between [6,800) and 7,000 kw*in less than or equal to 80.5 seconds and operates at
'b this load condition for at least 60 minutes.
This band is rneant as guidance to avoid routine overloading of diesel generator.
Loads in excess of the band or momentary variations due to changing bus loads shall not invalidate the test.
L
i l
TII 9Mit 8M80 l
AllACHMENT !!
l PAGE 18 Cf 52 g CTRICAL POWER SYSTEMS U
SURVEIL.ANCE REQUIREMENTS (Continued) b)
Verifying the diesel starts on the auto-start signal, energizes the emergency bustos with perm nently connected loads within 10 seconds, energizes the auto-connected emergency (accident) loads through the load sequencer and operates for greater than or equal to 5 minutes while its generator is loaded with the emergency loads.
After ener-gization, the steady state voltage and frequency of the emergency busses shall be maintained at 69002}90jvolts and 60 2 1.2 Hz during this test; and c)
Verifying that all automatic diesel generator trips, except engine overspeed and generator differential, are automatically bypassed upon IcJs of voltage on the emergency bus concurrent with a Safety Injection Actuation signal.
7)
Verifying the diesel generator operates for at least 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .
During the first 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months of this test, the diesel generator shal) be loaded to an indicated 7600 - 7700 k/ and during~ the
remaining 22 hours2.546296e-4 days 0.00611 hours 3.637566e-5 weeks 8.371e-6 months of this test, the diesel generator shall be loadedtoanindicated[6800]-7000kW.
The generator voltage and frequency shall be 6900 tf690] volts and 6011.2 Hz within 10 seconds after the start signaF; the steady-state generator voltage and frequency shall be maintained within these limits during this testf*Within 5 minutes af ter completing this l
24-hour test, perforia Specification 4.8.1.1.2/./)b);"
!D h 050 76 l
8)
Verifying that the auto-connected loads to each diesel
]
generator do not exceed the continuous rating of 7,000 kW; 9)
Verifying the diesel generator's capability to:
a)
Synchronize with the offsite cower source while the generator is loaded with its emergency loads upon a simulated restoration of of fsite power, b)
Transfer its loads to the offsite power source, and c) 8e restored to its standby status.
This band is meant as guidance to avoid routine overloading of the diesel generator.
Loads in excess of the band or momentary variations due to changing bus loads shall not invalidate the test.
"IfSpecification4.8.1.1.2f.6)b)isnotsatisfactorilycompleted,itisnot necessary to repeat the receding 24-hour test.
Instead, the diesel generator O
(
may be operated between 800]- 7000 kW for 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months or until operating tempera-ture has stabilized.
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III 9 Hit 68 124 AT)Atk*ENT11 PAGE260F$2 P. C. SOURCES w (
SURVEILLANCE REQUIREMENTS (Continued) b.
At least once per 92 days and within 7 days after a battery discharge with battery terminal voltage below 110 volts, or battery overcharge with battery terminal voltage above 150 volts, by verifying that:
1)
The parameters in Table 4.8-2 meet the Category 8 limits, 2:
There is no visible corrosion at either terminals or connectors, or the connection resistance of these items is less than 150 x 10 8 ohm, and 3)
The ave ge electrolyte temperature of 12 of connected cells is above[7 F.
l c.
At least once per 18 months by verifying that:
1)
The cells, cell plates, and battery racks show no visual indication of physical damage o.r. abnormal deterioration, 2)
The celloto-cell and terminal connections are clean, tight, and coated with anticorrosion material, 3)
The resistance of each cell-to-cell and terminal connection is
(
less than or equal to 150 x 10 8 ohm, and 4)
The battery charger will supply at least 300 amperes at 125 volts for at least 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .
d.
At least once per 18 months, during shutdown, by verifying that the battery capacity is adequate to supply and maintain in OPERA 8LE status all of the actual or simulated emerger.cy loads for the' design duty cycle when the battery is subjected to a battery service test; e.
At least once per 60 months, during shutdown, by verifying that the battery capacity is at least 80% of the manufacturer's rating when subjected to a perforinance discharge test. Once per 60-month interval this performance discharge test may be performed in lieu of the battery service test required by Specification 4.8.2.1d.; and f.
At least snce per 18 months, during shutdown, by giving performance discharge tests of battery capacity to any battery that shows signs of degradation or has reached 85% of the service life impected for the application.
Degradation is indicated when the battery capacity drops more than 10% of rated capacity fr its average on previous performance tests, or is below 90% of the manufacturer's rating,
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l COMANCHE PEAK - UNIf 1 3/4 8-13
IH*thMtf BB7$o ATIACHMENT !!
fAGE 27 CF $2 TA3LE 4.8-f H.
SATTERY SURVEILLANCE REQUIREMENTS k
CATEGORY ASI)
CATEGORY 8( )
PARAMETER LIMITS FOR EACH LIMITS FOR EACH ALLOWABLE (3)
DESIGNATED PILOT CONNECTED CELL VALUE FOR EACH CELL CONNECTED CELL Electrolyte
> Minimum level
> Minimum level Above top of Level indication mark,,
indication mark,
plates, and < \\" above and < \\" above and not maximum level maximum level overflowing indication mark indication mart Float Voltage 3 2.13 volts 1 2.13 volts (6)
> 2.07 volts Not more than 0.020 below the Spectfic
> 1.200(c) average of all
_ 1.195 connecte.d cel.ls Gravity Average of all Average of all connected cells connected cells
> 1.205 3,39$(5) 3
~
TABLE NOTATIONS (1) ForanyCstegoryAparameter(s)outsideteelimitis)shown,thebatteri may be considered CPERA8LE provided that within 14 hours1.62037e-4 days 0.00389 hours 2.314815e-5 weeks 5.327e-6 months all the Category 8 measyrements are taken and found to be within their allowable values, and provided all Category A and 8 parameter (s) are restored to within limits within the next 6 days.
(2) For any Category 8 parameter (s) outside the limit (s) shown, the battery may be considered OPERABLE provided that the category 8 parameters are within their allowable values and provided the Category 8 parameter (s) are restored to within limits within 7 days.
o (3) Any Category 8 parameter not within its allowable value indicates an inoperable battery.
(4) Corrected for electrolyte temperature (reference temperature of 77'F) and 1evel.
(5) Or batte n charging current is less than 2 amps when on charge.
(6) Corrected for average electrolyte temperature.
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CCMNCHE PEAK - UNIT 1 3/4 8-14
III 4Htt Eg /so ATTAD#ENT 11 isE M 0F $2 b
QNSITE POWER O!$Tn!8UTION l
LIN! TING CON 0! TION FOR OPERATION ACTION (Continued) b.
With one A.C. instrument bus or two A.C. instrument busses (consisting of one 7.5 KVA protection channel and one 1CKVA vital bus of tne same train) de-energized, re energize the A.C. instrument bus (ses) within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or be in at least HOT STAM08Y within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .
i With one A.C. instrument bus or two A.C. instrument busses (consisting c.
of one 7.5 KVA protection channel and e.10 KVA vital bus of the same 6).
train) operating with the associated inverter (s) not connected with__ (,)
the O.C. source (s), 's operating with the inverter 7not supplying the A.C. instrument bus (ou. with the instrument bus erergized from its i
associated bypass distriLution source), energize the A.C. instrument bus (ses) from its associated 0.C. bus within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or be in at t
least HOT STAN08Y within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within l
the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .
d.
With one O.C. bus not energized from its associated station battery, reenergize the O.C. bus from its associated station battery within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or be in at least HOT STAN08Y within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 20 hours2.314815e-4 days 0.00556 hours 3.306878e-5 weeks 7.61e-6 months .
SyRVE!LLANCE REQUIREMENTS J
4.8.3.1 The specified busses shall be determined energized in the required manner at least once per 7 days by verifying correct breaker alignment and 4
indicated voltage on the busses.
I 3
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CCMANCHE PEAK - UNIT 1 3/4 6-17
TH 96Hb B6760 ATTACHMENT 11 i
fAGE310F!2 ON5!TE POWER O!STRIBUTION SHUTOOWN s
LIMITING CON 01710N FOR OPERATION 3.8.3.2 As a minimum, the following electrical busses shall be energized in the specified manner:
One train of A.C. emergency busset, consisting of one 6900 volt and a.
two 480 volt A.C. er.' gency bus a,
t Two 118-volt A.C. instrument busses (channel-oriented) energized b.
from their associated inverters connected to their respective O.C.
busses; One train of A.C. Instrument busses consisting of two 118 volt A.C.
c.
instrument busses energized from their associated inverters connected to their respective O.C. busses.
Busses shall be of the same train as Specifications 3.8.J.2a. and d.; and d.
One train of 0.C. busses consisting..of.two 125-volt 0.C. busses.
energized from their associated battery banks.
Busses shall be of the same train as Specifications 3.8.3.2a. and c.
l APPLICABILITY MODES 5 and 6.
l
(
ACTION:
With any of the above required electrical busses not energized in the required manner, immediately suspend all operations involving CORE ALTERATIONS, posi'ive reactivity changes, or movement of irradiated fuel, initiate corrective acti e to energize the required electrical busses in the specified manner as soon as possible, and within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , d:;r:;wei-:: :nd v:nt th: "CG-thr::;h 04-leest.
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10 h M88 SURVE!LLANCE REQUIREMENTS 4.8.3.2 The specified busses shs11 be determined energized in the required manner at least once per 7 days by verifying correct breaker alignment and indicated voltage on the busses.
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COMANCHE PEAK - UNIT 1 3/4 8-18
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ELECTRICAL EQUIPMENT PROTECTIVE DEVICES CONTA! M NT PENETRAT!0N CONOUCTOR OVERCURRENT PROTICTIVE DEVICES L!MITING CON 0! TION FOR OPERATION 3.8.4 g M containment penetration conductor overcurrent protective devices l ' b' gh;r '- Tel: 3,5-1-shall be OPERABLE.
APPLICA8!LITY: MODES 1, 2, 3, and 4.
Oh*
ACTION:
With one or more of the containment penetration conductor overcurrent protective device (s) g';:n '- Tel: 3.S-1-inoperable:
ID 4: 0392.
J a.
Restore the protective device r ' :::r t = 2:r to OPERAPLE status or:
i' 1.
Deenergize the circuit (s) by racking out, locking open, or removing the inoperable ci n uit tr: d:r er protective device and p/ * * ") *' ~ " "declare the affected systee or component inopera s
trippingfth: :::: ht:d 5:0i;; ;ir;;it bn>hte within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .
.me<.. AJ p tni" the inoperable cin cit in d:r r protective device racked out, 3
J o =,
locked open, or removed at least once per 31 days thereafter; the p;gt, provisions of Specification 3.0.4 are not applicable to over-I g37g,nt' devices - "--" ' "--- in circuits which have their 5::ta; :icevit-tr;;h r: tripped.and their inoperable cir:d t y d W a i
,f beoeheee racked out, locked open, or removed; W '/%d d b e.
Ac.v ar i
N ' g /
2.
Deenergize the circuit (s) by tripping 4the ee***bted tre ICIC'!0
i ;;it tr;;i

r or racking out, locking open, or removing the P"'"" " \\r~syoperable.p4=uE and: within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , declare the af fected in 0""
stem or component in p.d6 M,,' f N eoeher to be tripped.gpe,ag e, and verify the
'f b
of he inoperab.le racked f......
\\
Ju'*
3 out, locked open, or removed at least once per 7 days thereafter; i
the provisions of Specification 3.0.4 are not applicable to
\\
overcurrent oevices in circuits which have their to % Otr44tt
--w 4 ;d:r:- tripped,pr their inoperable......... _.. racked out, g g, locked open, or removed; or y, J m** d* m l
b.
Be in at least HOT STAN08Y within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLO 7UTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .
SURVEh ~ cE REQUIREMENTS 4.8.44 M containment penetration conductor overcurrent protective devices l
- ' * ' shall be demonstrated OPERA 8LE:
a.
At least once per 18 months:
UhM 1)
By verifying that the medive voltage 6.9 kV and low voltage 480V i
switchgear circuit breakers are OPERABLE by selecting, on a rotating basis, at least one or 10% of the circuit breakers (whichever is greater) of each current rating and performing the cg.
l fc11 ewing:
f a)
A CHANNEL C/.LIBRATION of the associated protective rt: lays, l
b)
An integrated system functional test which includes simulated automatic actuation of the system and verifying that each
(
rt'ay and associated circuit breakers and control circuits i
function as cesigned, and CCMANCHE PEAK - UNIT 1 3/4 8-20 l
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III teMt 49Mo AtiAte(Ni 11 FAGE M CF st h
ELECTRICAL EQUIPMENT PROTECTIVE DEVICES w
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SURVE!LLANCE REQUIREMENTS (Continued) c)
For each circuit breaker found inoperable during these functional tests, one or an additional representative sample of at least 10% of all the circuit breakers of the inoperabt? type shall also be functionally tested until no more failures are found or all circuit breakers of that type have been functionally tested; and it
5 16, 2)
Iy selecting and functionally testing a representative sample of at least 10% of each type 480 V molded case circuit breakers and of lower voltage circuit breakers.
Circuit breakers selected for functional testing shall be selected on a rotating ~
basis.
Testing of these circuit breakers shall consist of
' injecting a current (
hw-ef-v.
3 gg 3 n _. m u..
.p ; p - ; _,;,, g g ;g,4,3;; g g3,
-'- ' d- -del:y trip;hr;;.0,- :nd : rtfying-that-the-4(esutt 19 ?. W to 4tT beuT&
-- + beeeker :;:r:tes-withie-the-ntee-deley-knd-idth-fee-that-
T
-: rr::t -opesif4+4-by-the-manwfeetweee -The-4*etant:n:::: -
r
.eleme*4-eh414-be-teeted by f*fecting-e-surr=t 0:29--to-e205-ef the-piehup-velwe-of-the-element-and-veef fying-th44-the-eleiw44 *
' tr=ter telpe-twete*446eevely-with-me-4ntentlenel-time-delay, Ed:d cooe-cieewit breaker-testing-shell also-fellow thn pr- :dar: ::::;t tMt g: :rel4y e - r:-tM". two-trip-eleseants.
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ti : del:y =d 100 tut===,
511 5: ' =h:d. tgCircuit
~ breakers found inoperable during functional testfrig'Shall be restored to CPERABLE status prior to resuming operation M or each circuit breaker found inoperable during these functional tests, an additional representative sample of 44 least 10% of all the circuit breakers of the ineperable type shall also be functionally tested untti no mere failures are found or all circuit breakers of that type P. ave been functionally tested;
~
b.
At least once per 60 menths by subjecting each circuit breaker to an inspection and preventive maintenance in accordance with procedures prepared in conjunction with its manuf acturer's recomunendations, s
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,m o u r t
TAat.E 3.8 1 (Continued)
CONTAINMENT PENETRATION CON 00CTOR QVERCURRENT PROTECTIVE,0EVICES DEVICE MBER SYSTEM AND LOCA7!0N METRTO
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125V 0 LightinV Emergency DC Lighting a.
Pri ry Breaker SREAKElhl0 CAT!0N G. E. SKR.
AND NUMB H TYPE DC Pane 1bo\\
ard TFJ 102-1, Ckt 2
b.
Backup Device - /A (7use) 7.
125V DC Control Power Various a.
Primary Cevices - N/A uses I
b.
Backup Breakers GENERAL ELECTRIC
-tA6 4 -
PANEL 80ARD N0.
C C. No.
SREAKER TYPE l
' XE01-1 6,7,8,9,10 TED l
-OF XE02-1 6,7,8,9,10 TED 03-XD2-3 4,
12,14,17 TED 04-
f91-1 li 14,9r10 TEG-M 1E02-1 7,10 12,15,16,17 TED M-
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l COMANCHE PEAK - UNIT 1 3/4 8-34 1
f111N+vF88180 AllACM(Ni 13 PA E 6 0F 8 1PECIALTEstEXCEPTIONS 3/4.10.4 REACTOR C00LANT LOOPS LIMITING CON 0! TION FOR OPERATION i
3.10.4 The limitations of Specification 3.4.1.2 nay be suspended during the performance of hot rod drop time measurements in M00E 3 provided at least two reactor coolant loops as listed in Specification 3.4.1.2 are OPENA8LE.
APPLICA8!L!rY:
During performance of hot roa drop time measurements.
ACTION:
With less than the above required reactor coolant loops OPERA 4LE during the performance of hot rod drop time measurements, immediately open the reactor t.'ip breakers and comply with the provision of the action statements of Specification 3.4.1.2.
s.
SURVE!LLANCE REQUIREMENTS
([
4.10.4 At least the above rsquired reacter coolant loops shall be determined OPERA 8LE within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months prior to the initiation of hot. rod drop time measure-ments by verifying :;rr;rt kreaker alignments and indivited power availability (b.
and by verifying the indicate secondary side water level to be greater than or equal to 105 narrow range.
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ixx19ttt u lto AliaQuei13 PAGE 8 0F 1 3/4.10 $PECIAL TEST EXCEPTIONS BASES 3/4.10.1 SHUT 00wN MAR 13 This special test exception provides that a minimum amount of control rod worth is immediately available for reactivity control when tests are performed for control rod worth nessurement.
This special test oception is required to permit the periodic verification of the actual versus predicted core reactivity condition occurring as a result of fuel burnup or fuel cycling operations.
3/4.10.2 _ GROUP HEIGHT. INSERTION. AND POWER O!$TRIBUTION LIMITS This special test exception pers.its individual control rods to be post-tiened outside of their normal group heights and insertion limits during the performance of such PHYSICS TESTS as those required to:
(1) measure control rod worth, and (2) determine the reactor stability index and damping factor under menon oscillation conditions.
~
3/4.10.3 PHYSICSTEST)
This special test exception permits PHYSICS TESTS to be performed at loss th.n or equal to 5% of RATED THERMAL POWER with the RCS T,,, slightly lower k
than normally allowed so that the fundamental nuclent characteristics of the core and related instrumentation can be verified.
In order for various char-acteristics to be accurately measured, it is at times necessary to operate outside the normal restrictions of these Technical Specifications.
For instance, to measure the moderator teacerature coefficient at BOL, it is necessary 4e. position the various control rods at heights which may not l
normally be allowed by Specification 3.1.3.6 and the RCS T,,, may fall slightly below the minimum temperature of Specifiestion 3.1.1.4.
3/4.10.4 #EACTOR COOLANT LOOPS This special test exception is required to perform certain STARTUP and PHY51CS TESTS under no flow conditions.
3/4.10.5 PO$! TION INDICATION SYSTEM - SHUT 00wN This special test exception permits the Digital Rod Position Indicator (s) to be inoperable during rod drop time measurements.
The exception is required since the data necessary to determine the rod drop time a're derived from the induced voltage in the position indicator coils as the red is dropped.
This induced voltage is small compared to the normal volta be observed if the Digital Rod Position Indicator (s) ge and, therefore, cannot remain OPERAILE, The exception to the requirement for the Digital Red Position Indicator to be OPERA 8LE during the withdrawal of the rods for the initial calibration of the position indication system is required because the OPERA 8ILITY of the Digital
(
Red Position Indication System can only be determined by withdrawing the control rod.
The limitation on Keff during this evolution provides the necessary assurance that inadvertent 4+iti d ly will be avoided.
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III4eHt 68760 ATTACle(Ilf 14 PAGE200FJO RA0!OACT!VE EFFLUENTS EXPLO5!VE GAS MIXTURE LIMITING CON 0! TION FOR OPERATION 4
3.11.2./ The concentration of oxygen in the WA57E GAS HOLOUP SYSTEM shall be 3-limited to less than or equal to 32 by volume whenever the hydrogen concentration exceeds 45 by volume.
APPLICA8!LITY:
At all times.
/CTION:
a.
With the concentration c ' oxygen in the WASTE GAS H0 LOUP SYSTE4 greater than 35 by volume but less thrn or equal to 4% by volume, reduce the oxygen concentration to the above limits within 44 hours5.092593e-4 days 0.0122 hours 7.275132e-5 weeks 1.6742e-5 months , b.
With the concentration of cxygen in the WASTE GAS HOLDUP SYSTEM greater than 45 by volume and the hydrogen cor:entration greater than 45 by volume, immediately suspend.all additions of waste gases.
to the system and rnduce the concentration of oxygen to less than or squal to 45 by volume, then take ACTION a,,
above.
c.
The provisions of specificationu 3.0.3 and 3.0.4 are not applicable, l
l SURVE!LLANCE REQUIREMENTS 4.11.2.fTheconcentrationsofhydrogenandoxygenintheWASTEGA5HOLOUP l#
SYSTEM shall be determined to be within the above limits by se-d r e!s 9,,l]g monitoring the waste gases in the WASTE GAS HOLOUP SYSTEM with the hydrogen and oxygen monitors required CPERA4LE by Table 3.3-X of Specification jO' 3.3.3.X.
10 1
COMANCHE PEAK - UNIT 1 3/4 11-15
rtI-tolto cnincanster is CpSES Technical Specifications
?nst 2A or to NRC Draft 2 Harkup Section 3/4.12 Change 108 Justification for Change 0440 Action "b" is being modified to take into account unavailability of samples in this region.
For certain samples, i.e. milk, vegetation, etc., collection of these samples may not be readily possible depending on where the new sampling location is.
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3/4.12 RA0!0 LOGICAL ENVIRONMENTAL MONITORING
,.D W l1 3/4.12.1,_ MON!TORING PROGRAM j
RELOCME i,, Oa LIM! TING CO'1! TION FOR OPERATION The Radiological Environmental Monitoring Program shall be! conducted I
3.12.1 as specified'in Table 3.121.
s r
APPLICABILITY:
At all times.
ACT!0N:
With the Radiological Environmental Monitori g Program not being a.
f conducted as specified in Table 3.12 1, precare and submit to the Cosimission, in the Annual Radiologicaf Environmental Operating Report reovired by Specification 6.9.1.)' a description of the reasons fo? not conducting the program as req (red and the plans for preyenting
~
eM 18 1: 0431 b.
With the. level of radioactivity as' the result of plant of fluents in an environmental sampling mediuvat.a.specified locattor exceeding the reporting levels of Table 1.12-2 when averaged over any calendar quarter, prepare and submit,16 the Commission within 30 days, pursuant to Specification 6.9.2, a special Report that identifies the cause(s)
)
for exceeding the limit (s7 and defines the corrective actions to be taken tc reduce radioactive effluents so that the potential annual dese" to a MEM8ER OF TFE PUSLIC is less than the calendar yeai limits of Specifications 3.11.1.2, 3.11.2.2, or 3.11.2.3, Wne.1 more than one of the radionugtides in Table 3.12-2 are detected in the sampling medium, this report shall be submitted if:
/
eencentration (11 concentration (2) 1 report level C,)
repor',ing level (2)
A 1*O l
When radienuclides other than those in Table 3.12-2 are detected and i
are the,pesult of plant effluents, this report shall be submitted if the potential annual dose
to a MEMBER OF THE PUBLIC from all radio-i nuc1) des is equal to or greater than the calendar year slimits of Specification 3.11.1.2, 3.11.2.2, or 3.11.2.3.
This report is not required if tM measured level of radioactivity was not the result
[4fplanteffluents;nowvver,insuchanevent,theco9ditionshall
)
be reported and described in the Annual Radiological Environa ntal Cperating Report required by Specification 6.9.1.E.
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"The setnodology and parameters used to estimate the potential annual dose to N I
a MEM8ER OF THE PUBLIC shall be indicateo 1.s this report, A e r l i, m.3 o < 3 rc 4.n a.by.k -de, e c '. f, #,.\\, A 4.g e4M c c
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f TARtE 3.12-1 (Coatinued)
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on RADIOtOGICAL f NVIRODMENTAL MONITORING PROGRAM
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E Rf PR[SENI ATIVE O$
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[MPOSURE PATHWAY 5AMPLIS AND Mt M M ME m F$C gg) g g
AND/OR SAMrif SAMPt[ LOCATIOtts C0ttECIION FRfQUINCY Of ANAtYSIS 4.
Ingestion a.
Milk Samples from available milking Seelmonthly when Gamma isotopic nd ca animals in three locations animals are on pasturo; I-131 analysis semi-within 5 km distance having the monthly at other L M s.
monthly when nimals highest dose potential.
If there are on pasture; monthly are none.-then one sample from at other times.
available allking animals in each of three areas between 5 to 8 km distant where dos'esgpre calculated 4
d.
Y to tse greater than T"Fr\\ per P iP -1 em
/
yr. ( 10) One sample f ree milking a-g animals at a co.. trol local.i6n l5 s
to 30 km distan. 4:ni in the lea prevalent wind direc/' ion. ( 3 )
/
isotopic analysis (5) b.
Iish and One sample of-each commercially Sample semiannually.
ha Inverte-and recrestionally importact on edible portions.
brates specic Vin vicinity of plant digharge area.
/One sample of same species in areas not intluenced by plant y
discharge.
c.
food One sample of each principal At time of harvest (II).
Isotopic analyses (5)
- Products class of food products free on edjble portion.
any area that is irrigated by water in which liquid plant wastes have been discharged.
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TII WHf 808c ATTACHMENI 15 fASE 16 0F 20 DRAiT RA0!0 LOGICAL ENv!RONMENTAt. MOMITORING MOCME 3 J.12. 2 L.ANO USE CENSUS
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LIMITING CON 0! TION FOR OPERATION 3.12.2 Land Use Census shall be conducted and shall identify within a distanceo(0km(5 miles)thelocationineachofthe16 meteorological sectors of a nearest milk animal, the nearest residence, and the nearest garden
of g ster than 50 m8 8
(500 f t ) producing broad leaf vegetation.
APPLICABILITY:
t all times.
ACTION:
a.
With a Landsyse Census identifying a location (s) that yields a calculated dase or dose commitment greater than the values currently being calculated in $pecification 4.11.2.3, pu*suant to Specifica-tion 6.9.1.4, thentify the new location (s) in the next 3emiannual Radioactive Eff1 nt Release Aaport.
b.
With a Land Use con vs identifying ( 14 cation (s) that yields a
~
calculated dose or dote commitment (via the same exposure pathway) 20% greater than at a tocation from which samples are currently being obtained in accord'acce with Specification 3.12.1, add the new k
location (s) within 30 da>A to the Aadiological Environmental Mont-tu toring Program given in the'00 cms 4 The sampling location (s), exclud-L gn
(
Ing the control station locat n, having the lowest calculated dose or dose commitment (s), via th ame exposure oathway, may be deleted dp "E from this monitoring program af tgr October 31 of the year in which j
WL***
tMs Land Use Census was conducteel, Pursuant to Specification 6.14, submit in the next Semiannual Radtbective Effluent Release keport documentation for a change in the 00CM includi.99 a revised figure (s) and table (s) for the 00CM reflecting t e new location (s) with,informa-tion supporting the change in sampling orations.
c.
The provisions of Specifications 3.0.3 a 3.0.4 are not applicable.
"Broad leaf vegetation sampling of at least three different inds of vegetation may be performed at the EXCLU510N AREA 8NNOARY in each of t% different direction sectors with the highest predicted 0/Qs in lieu of the garden census.
Specifications for broad leaf vegetatie.1 sacoling in Table 3.12-1. Part 4.c.,
shall be followed, including analysis of control saeples.
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CCKANCNE PEAK - UNIT 1 3/4 12 16
TIX 909tf tt1Bo Af f 6 chm (NT 15 FAGE 11 W 20
.00M U.5-RA010LOGICALTWIRONMENTAL_ MON!TORING w
-x fH 0441
$URV(!LLANCE R(QUIR[M(MS_
i
,n 4.12.2 The Land Use Census s ondugte ring the growl 12 season at I
least once per 12 months using tha ga*,fonthatwillprovidethebest results, such as by a door-to doo[r e evey, serial survey, or by consultingl 1
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included in the Annual R4414Togical Environmental De ating Report pursuant, to 5.*ecification 6.9.1J/
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' RADIOLOGICAL (NV!RO W(NTAL MON!TORING p
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3/d.12.3 INT [RLASORATORY CcMPAR!s0N PROGRAM bk 10 h M LIMITiWGCCA0!TIONFOROPERAT!cN
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3.12.3 Analyses shall be performed on all radioactive materiais, sapallea as part of an Inter 14boratory Comoalison Program that has been approved by the Commission, that correspond to simples required by Table 3,12 1.
/
APPLICAI!LITY: 'At all times.
ACTION:
a.
With analyses not Seing performed as required above, report the corrective actions taken to prevent &' recurrence to the Commission in the Annual Radiological Environmental Operating Report pu.'suant i
to Specification,6.9.1.3.
b.
The provisions of 5pecificati 3.0.3 and 3.0.4 are not applica01e.
$URVE!LLANCE R(QUIREM(NTS 4.12.3 The Interlaboratory Compar s.on Program shall be described in the 00CM.
A summary of the results obtained as'part of the above required Interlaboratory i
Comparison Program shall be irwliluded (q the Annual Radiological Environmental
[
s t
Operating Repcrt pursuant tylpecification %9.14
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TXX-99tt216160 ATIACHMENT 16 s
PAGE 10 OF 12 DESIGN FEATURES
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5. 3 REACTOR CORE FUEL ASSEM8 LIES 5.3.1 The core shall contain 193 fuel assemblies with each fuel assembly containing 264 fuel rods clad with Zircaloy-4 except that limited substitution of fuel rods by filler rods (consisting of Zircaloy-4 or stainless steel) or by vacancies may be made if justified by a cycle specific reload analysis.
Each fuel rod shall have a nominal active fuel length of 144 inches.
The initial core loading shall have a maximum enrichment not to exceed 3.15 weight percent U-235.
Reload fuel shall be similar in physical design t th initial core loading and shall have a maximum enrichment not to exceed 3.5 weight percent U-235.
CONTROL R00 ASSEMBLIES 5.3.2 The core shall contain 53 full-length control rod assemblies.
The full-length control rod assemblies shall contain a nominal 142 inches of absorber material.
The nominal values of absorber material shall be 95.5%
hafnium with the remainder zirconium.
All control rods shall be clad with stainless steel tubing.
5.4 REACTOR COOLANT SYSTEM DESIGN PRESSURE AND TEPOERATURE 5.4.1 The Reactor Coolant System is designed and shall be maintained:
InaccordancewiththeCoderequirementsspecifiedinSection.[5.2/
a.
of the FSAR, with allowance for normal degradation pursuant to the applicable Surveillsnee Requirements.
b.
For a pressure of 2,485 psig, and c.
For a temprature of 650*F, except for the pressurizer which is 680*F.
VOLUME 5.4.2 The total water and steam volume of the Reactor Coolant System is lh l3F [12.500]3100 cubic feet at a nominal T of 589.L*F.
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5. 5 METEOR 010GICAL TOWER LOCATION
5. 5.1 The primary meteorological tower sha11 be located as shown on rigure b.1-1.
COMANCHE PEAK - UNIT 1 5-5
III 8 Hit e,80 AliACHM[NTit' PME 4 of 41 CPSES Technical Specifications NRC Draft 2 Markup Section 6 (Cont'd)
Change IDd Justification For Change 0531 Added the requirements to perform an audit of the Radioactive Effluent and Environn.cntal Monitoring Manual (REEMM).
The REEMM incorporates ti.e Radiological Environmental Monitoring Program, therefore due to the increase in the material to be audited the frequency has been increas,ed to 24 months.
This also makes it consistent with the 00CM.
This change is similar to that licensed at Millstone 3.
QWy@g{405 @, [ 63' requirement to perform an audit of the Technical Specificatio vement Pro ram (TS!P).
This is to ynge SA op 4g ensure that all requir ocated to the TSIP are cgnqqq 2a,+F os34 implemented in a correct manner.
0536 Added the Station Operations Review Committee for 0314 distribution of Licensee Event Reports and made minor editorial changes to more closely agree with the standard technical specifications.
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TIX-00Hf 85180 ATTACHMENT 17
" # " '3 ',
CPSES Technical Specifications NRC Oraft 2 Markup Section 6 (Cont'd)
Change ID#
Justification For Change 0907 These changes remove the requirement to maintain an 0916 organizational chart in the Tech Specs.
The information that is being removed from the Tech Specs is contained in the FSAR and updated in accordance with 10CFR50.71.
This change is made in confortnance with Generic Letter 88-06.
0908 NRC initiated these changes during the 4/12/88 meeting 0921 with OSP and NRR.
0909 This note is for, and included on the previous page, therefore should be deleted.
0911 7'
ange deletes the word "may."
With the word "may" sowesmb y wx ee78o include ld r l;uire that 50RC review all proposed M4-QUA*0NT M
?AGG SA c5 procedures, chang rocedures, equipment, systems.
C4MGs Ib *41 facilities, proposed tes x eriments.
This would be o9 tl 50 time consuming that it would s icantly detract from the real responsibility of 50RC to be co t of actual safety concerns (F
0912 Deleted the requirement for 50RC to recommend in writing approval or disapproval of 6.5.1.61 through 6.5.2.61.
This is based on the Standard Technical Specifications only rewire that responsibilities.from 6.5.1.6a through 6.5.1.6e have this specific requirement.
Each item (i) through (k) have separate requirements, 6.13 through 6.16, and (1) already specify where to forward tile required reports.
l 4
0
ATTACMENT 17 FAGE 8 CF 41 CPSES Technical Specifications
~
NRC Draft 2 Markup Section 6 (Cont'd) '
Change 10#
Jus *ification For Change 0913 See 10# 0527 0914 See 10# 0536 0916 See 10f 0907 0917 The change to training requirements is made based cn the revision to 10CFR55 and NUREG-1262.
The reference to Appendix "A" to-10CFR55 has been deleted since it no longer exists.
The reference to the H.R. Denton letter dated March 28, 1980 is removed due to the guidance put out in NUREG-1262.
The rule supersedes and includes the requirements of the Harold Denton letter of March 28, 1980 (Q/A number 385 of NUREG-1262).
The only requirement to be satisfied is to meet the minimum requirement of the revised rulc.
0920 e quorum requirement to 5 members and the sonesaud M Tvx-8878o chairman vic ers arid the chairman.
This is based agcktjT on having the majority ORC areas represented in us n't hDe n Ir o91o every 50RC meeting.
0921 See 108 0908 O
e L
TkV-6878o amtgmg a CPSES Technical Specifications pass my 4 NRC Dratt 2 Markup Section 6.0 Change 10#
Justification for Change 0534 Added the requirement to perform an addit cf the Technical Specification Improvement Program (TSIP).
This is to ensure that all requirements located to the TSIP tre implemented in a correct manner. A 24 morth audit frequency was chosen based on ANSI N18.7-1976/ANS-3.2, Section 4.5, Audit Program, as modified by Paragraph C of l
Reg. Guide 1.33, which requires in part that "Audits of selected aspects of operational phase activities shall be performed with a L
frequency commensurate with their safety significance and in such a manner as to assure that an audit of all safety-related functions is completad within a period of two years... Conforaiance of facility operations to provisions contained within the Technical Specifications and License Conditions - at least once per 12 months."
The requirements that TV Electric is proposing to relocate from the Technical Specifications to the Technical Specification Improvement Program (TSIP) are not of the l
same safety significance as the requirements that are contained in the Technical Specifications.
Therefore, the proposed two year audit period for activities governed by the TSIP is consistent with the intent and regulatory position of ANSI N18.7-1976/ANS-3.2 and Reg. Guide 1.33.
0911 These review requirements are not in the standard I
technical specification and add no enhancement to safety.
Under 6.5.1.6b, SORC is required to review all safety evaluations completed under the provisions of 10CFR50,59 to verify that such actions did not constitute an unreviewed safety question.
In reviewing of the safety evaluation it is necessary to review the document or changes to the document in order to make a rational decision on whether it constitutes an unreviewed safety question or not.
This change will eliminate confusion as to the interpretation of 6.5.1.6c and will make the CPSES Technical Specifications more consistent with the standard technical specifications.
This justification supersedes the justification for 10 #0911 in TU Electric Letter TXX-88512.
0920 CPSES inadvertently made the correlation of the number of i
areas represented to the number of members on 50RC.
The fact is that a given member of 50RC may in fact be representing more than one area on SORC. This change is to prevent CPSES from being overly restrictive and return to the standard technical specification wording.
This i
justification supersedes the justification for 10 #0920 in TV Electric Letter TXX-88512.
M Tu - 8676o AlfACHme uT II CPSES Technical Specifications 7nu 15oF 41 NRC Draft 2 Markup Section 6.0 (Cont'd) t Change 10#
Justification for Change
'5127 Added the ANSI N18.1-1971 Section 4.4 to the experience requirements of 50RC sitice this is the section that specifies the experience requirements for some of the-managers that will be members of ! ORC.
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ATTACHMENT 11 PAE la # 41 INSERT FOR NEW PAGE G-5 (OLD 6-7)
SECTION 6.5.1.2 ngw 6.5.1.2 The SORC shall be composed of managers or individuals reporting directly to managers from the rb4 Tit 7 areas listed below and meet the requirements of ANSI N18.1-1971 Section 4.2 4'for required experience.
or 4A operations Maintenance Instrumentation and Controls Technical Support Radiation Protection Quality Assurance Emergency Planning Security Testing The Manager, Plant operations shall serve as the chairman of f-SORC.
A senior health physicist is acceptable for the Radiation Protection representative on SORC.
The SORC members shall be designated, in writing, by the Vice President, Nuclear Operations.
S 4
(
L
m68878o i
ATTACHMENTIF PAGE 19 0F 41
)
ADMINISTRATIVE CONTROLS A
MEETING FREQUENCY
6. 5.1. 4 The 50RC sna11 meet at least once per calendar month and as convened by the 50RC Chairman or his designated alternate.
QUORUM
6. 5.1. 5 The quorum of the 50RC necessary for the performance of the 50RC
??
responsibility and authority provisions of these Technical Specifications shall consist of the Chairman or his designated alternate and Jewe senters including alternates, c';; y 9.(g)
[ 4 P ^
U RESPONS!8!LITIES p m.J w e, is e. = n
J d Irs 6.5.1.6 The 50RC shall be responsible for:
p App M r A ef A e *4*"7 t
K Cawias1,11, b'.C m 2 Tear w a.
Review of :!' St:t!:n 5 f-!:tr:tf= "r:= durn; mg 10 8: 0522 b.
Review of the safety evaluations fort (1) procedures, (2) change -
to procedures, equipment, systems or facilities, and (3) tests or experiments completed under the provision of 10 CFR 50.59 to verify that such actions did not constitute an unreviewed safety question:
c.
R:vi Of ;r:;;;;d pren der;: =d ch=;n :: ; nder;;, :;;ipment,
y:t=: cr feeilitie: which =-, involve ;n wrirevi;w;d ;;fety ques-
-ti= :: def4n;d in 10 CF" 50.5" ;r involve: ; ch=;; in Technteet-gg
-Sew 4+4eati:n:;
;_ g;;
-d.
" v f r.: Of pr;;;;;d test er =;;rkents-whfen =, invciv; en i
f
-uncev ewed-sa ety-tweetion a; d;ff a;d in 10 Cf" 50 5" or r;qufres-e-thenge-in-Teehn te el-Spee M4e a t fenet c g.
Review of proposed changes to Technical Specifications or the Operating License; d t.
Investigation of all violations of the Technical Specifications including the forwarding of reports covering evaluation and recom-sendations to prevent recurrence to the vice President Nuclear Operations and to the ORC; e ).
Review of reports of operating abnormalities, deviations from ex-pected performance of plant equipment and of unanticipated defici-encies in thi design or operation of structures, s stems or j
components that affect nuclear safety; h.
Review of all REPORTA8LE EVENTS;
<} \\.
Review of the Security Plan and shall submit recommended changes to l(
the CRC; l
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COMANCHE PEAK - UNIT 1 6-\\ f.
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AllACMENT 11 PAGE 20 of 41 g %.g a n.x.- w tae ADMIN!$TRATIVE CONTROLS RESPONS!8!LITIES (Continued) h (.
Review of the Emergency Plan and shall submit recommended changes to t e 0RC;
,,,,,w,,,,,,J,a,,,,,,,,,,,,_,,,,,,,,,.
6 K.
Review of changes to the PROCESS CONTROL PROGRAM,A 0FFS!TE 00SE CALCULATION MANUAL, and Radweste Treatment Systems; ID 1: 0523
'1 Review of any accidental, unplanned or uncontrolled radioactive
)
release including the preparation of reports covering evaluation, recommendations, and disposition of the corrective action to prevent recurrence and the forwarding of these reports to the Vice President, Nuclear Operations, and to the ORC; L X.
Review of Unit operations to detect potential hazards to nuclear safety; e i K.
Investigations or analysis of special subjects as requested b the Chairman of the CRC or the Vice President, Nuclear Operatio
a. [
Review of the Fire Protection Program and revisions thereto; a J
6. 5.1. 7 The 50RC shall:
h 10 h OH3 10 h 0527 (ses s ssL% 6 ')
sulyJ4 rnana.$ <.e-e a.
Reconsend in writing to the Yi:: "r::f d::t, " :!:: 5:::tt::: approval l
or disapproval of itses considered under Specification 6.5.1.6a.
through g 5, j, t, :nd ' 2:::, prior to their implementation; O-l c
10 h 0912 I
b.
Render determinations in writing with regard t' whether or not each ites considered under Specification 6.5.1.64. through g. :nd :.
constitutes an unreviewed safety question; and d
Provide written notification within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months to the Executive Vice c.
President Nuclear Engineering and Operations and the Operations Review Coenittee of disagreeeent between the 50RC and the 44ee-
r;;ff: t, 5 :!::r 5 :r:t!:::; howevsr, the Vice President, Nuclear eb.
s 3 Operations shall have responsibility for resolution of such dis-agreements pursuant to Specification 6.1.1.
RECOROS
~
^ mqar ft ID h 0529 h
6.5.1.8 The 50RC shall maintain written sinute's of each 50RC meeting that, at a sinimum, document the results of all 50RC activitiae performed under the i
responsibility provisions of these Technical 3pecifications. Copies shall be provided to the Vice President-Nuclear Operations and the Operations Review Connittee, 9.tasa.5 A m u isg,.;5;,. 4 y,,p,,,,,,,3 p,,7,
p.
M "tv4h useet.
r l
1D h U$M i
CCMANCHE PEAK - UNIT 1 6-\\7 l
- - - - ~ - - - -
~
TIX-999tt g e eo AVIACMENT17 PAGE350F41 ADMIN!'5TRAT!VE CONTROLS RECOR05
~
6.5.2.9 Records of ORC activities shall be prepared, approved, and distribu-ted ar, Indicated below:
Minutes of each ORC seeting shall be prepared, approved, eM fa. -
a.
warded to the Vice President, Nuclear Operations and Executive Vice President, Nuclear Engineering and Operations within 14 days following each meeting; i
b.
Reports of reviews encompassed by Specification 6.5.2.7 shall be prepared, approved, and forwarded to the Vice President Nuclear Operations and Executive Vice President Nuclear Engineering and Operations within 14 days following comp,letion of the review; and Audit reports encompassed by Specification 6.5.2.8 thall be for-c.
warded to the Vice President, Nuclear Operations a i Executive Vice President, Nuclear Engineering and Operations and to the management positions responsible for the areas audited within 30 days after completion of the audit by the auditing organization.
J. 5. 3 TECHNICAL REV!EV AND CONTROLS 6.5.3.1 Activities which affect nuclear safety shall be conducted as follows:
/
Procedures required by Specification 0.8 and other procedurer which a.
af fact plant nuclear safety, and changes thereto, shall be prepared, reviewed and approved.
Each such procedure or procedure change shall be reviewed by a qualified individual /groupenther than the e
I
' ~tndividual/ group which prepared the procedure or procedure change, but who say be froe the same organization as the individual /grouD 10 h 09C$
which prepared the procedure or procedure change. The Vice Presi-l dent, Nuclear Operations, shall approve Station Administrative Procedures, Security Plan Implementing Procedures, and Emergency Plan !aplementing Procedures. Other precedures shall be approved by h eanager, the g;%ri A:;;r:; ? =t' Mt;, = 2:f;;r:ted -by the Vice Presi-ED.
as previously -- cent, nuclear operations, in writtig.
Individuals responsible for designated procedure reviews shall be seabers of.the Nuclear Operations Staff
~
previously designated by the Vice President Nuclear Operations.
Changes ta procedures which de not change the intent of approved procedures may be approved for isolementation by two seabers of the
^
Nuclear Operations Staff, at least one of whom holds a Senior I
Operatar License, provided such approval is prior to implementation and is documented.
Such changes shall be approved by the original
)
approval authority within 14 days of implementation; b.
Proposed tests and experfeents which affect plant nuclear safety Ord r; n:t :::r;;;;d ' th: " nd S h t, 'n:Pj:t: ";;;rt r 7.ui;e! ;;;;f t;e;iew shall be prepared, reviewed, and
/
eeting the experience requirements of ANSI N18.1-1971, Section3 4'2' 4'3' 4
4.5.1 (Licensed Operators), 4.5.2, or 4.6 CCMANCHE PEAK - UNIT 1 6-M ll
III etHM 80700 ATTACEENT17 PAG ( 26 07 41 a
l)
.o.
ADMINiSTRAT!VE CONTROLS TECHNICAL REVIEW AMO CONTROLS (Continued)
Each such test or experiment shall be reviewed by a approved.
~
( qualified individual /arovoaother than the individual / group wnich
{pg 09gg
/ prepared the proposeo test or experiment.
Proposed test and exceri-I t
ments shall be approved before impleeentation by the Manager, Plant Operations.
Individuals responsible for conducting such reviews shall be seeCers of the Nuclear Operations Staff previously designated by the. Vice President, Nuclear Operations; Proposed changes or modifications to pl?nt nWiear safety related c.
structures, systems anc components shall be reviewed as designated by the Vice President, Engineering and construction.
Each sucn sodification shall be reviewed by a qualified individual / group M
1:30903 m other than the individual / group which designed the modification.
but who say be from the same organization as the individual / group l
which designed the modifications.
Individuals / groups responsible for conducting such reviews shall be previously designated by the Vice President, Engineering and Construction. Proposed modifica-tions to plant nuclear safety-related structures, systems and components shall be approved by the Manager, Plant Operations prior to implementation; Each review conducted in accordance"with the requirements oI5peci-d.
fications 6.5.3.la, 6.5.3.lb, and 6.5.3.lc, shall include a deter-aination of whether or not additional cross disciplinary review is necessary.
If deemed necessary, such review shall be done in accordance with the appropriate qualification requirements; Each review shall include a determination of whether or_nt_an e.
unreviewed safety question is involved.
MNRC approvalj W
8"-'"--*
I'e r
-o* Items involving unreviewed safety questions,% hall be obtained prior to the Manager, Plant Operations, approval for f aplementation; and f.
The Security Plan and Emergency Plan, and isolementing procedures, shall be reviewed at least once per 12 months.
Recommended changes to the teclamenting procedures shall be approved by the " :;a,
l
":nt %:- ti:n:.
Recommended changes to the Plans shall be I
o reviewed pursuant to the requirements of Specifications 6.5.1.6 and a
6.5.2.4 and approved by the "r:;;r, "Unt O:::ti: :.
NRC approval 2
{
shall be odtained as appropriate.
t.vm hu Ay,u.
og,twi 6.5.3.2 Records of the above activities described in 6.5.3.1 shall be provided to the Vice President, Nuclear Operations, 50RC, and/or ORC as necessary for required reviews.
6. 6 REPORTA8(I EVENT ACTION 6.6.1 The following actions shall be taken for REPORTA2LE EVENTS:
The Commission shall be notified and a report submitted pursuant to a.
the requirements of 10 CFR 50.73 and b.
Each REPORTA4LE EVENT shall be reviewed by the SCRC, and the results of this review shall be submitted to the CRC and the Vice President Nuclear Operations.
{meetingtheexperiencerequirementsofANSIN18.1-1971, Sections 4.2, 4.3, 4.4, or 4.6 i
COMANCHE PEAK UNIT 1 6 K l1 i
meeting the experience requirements of ANSI N18.1-1971, Section 4.6 j
INSTRUCTION SHEET The following attachment to TV Electric Letter TXX-88780 is intended to update and in some cases supersede certain pages from a previous TV Electric Letter, TXX-88512 dated June 23, 1988.
The following instructional information is provided to assist in this update.
REMOVE (from TXX-88512)
/NSERT (from TXX-88780)
ATTACHMENT 1 Page 2 of 13 Page 2 of 13 Page 2A of 13 Page 5 of 13 Page 5 of 13 ATTACHMENT 2 Page 2A of 20 Page 11 of 20 Page 11 of 20 Page 19 of 20 Page 19 of 20 ATTACHMENT 3 Page 2A of 12 Page 6 of 12 Page 6 of 12 Page 6A of 12 Page 6B of 12 Page 6C of 12 Page 6D of 12 Page 6E of 12 Page 6F of 12 Page 6G of 12 Page 10 of 12 Page 10 of 12 Page 12 of 12 Page 12 of 12 Page 12A of 12 ATTACHMENT 4 Page 3A of 29 Page 5 of 29 Page 5 of 29 Page 12 of 29 Page 12 of 29 Page 13 of 29 Page 13 of 29 Page 14 of 29 Page 14 of 29 Page 27 of 29 Page 27 of 29 Page 28 of 29 Page 28 of 29 ATTACHMENT 5 Page 4A of 25 Page 5 of 25 Page 5 of 25 Page 14 of 75 Page 14 of 25
r-INSTRUCTION SHEET Page 2 REMOVE (from TXX-88512)
INSERT (from TXX-88780)
ATTACHMENT 6 Page 9A of 105 Page 10 of 105 Page 10 of 105 Page 17 of 105 Page 17 of 105 Page 18 of 105 Page 18 of 105 Page 22 of 105 Page 22 of 105 Page 30A of 105 Page 31 of 105 Page 31 of 105 Page 41 of 105 Page 41 of 105 Page 46 of 105 Page 46 of 105 Page 60 of 105 Page 60 of 105 Page 62 of 105 Page 62 of 105 Page 68A of 105 Page 72 of 105 Page 72 of 105 Page 93 of 105 Page 93 of 105 Page 98 of 105 Page 98 of 105 Page 99 of 105 Page 99 of 105 Page 100 of 105 Page 100 of 105 Page 102 of 105 Page 102 of 105 Page 104 of 105 Page 104 of 105 ATTACHMENT 7 Page 6 of 66 Page 6 of 66 Page 7 of 66 Page 7 of 66 Dage 7A of 66 Page 20 of 66 Page 20 of 66 Page 27 of 66 Page 27 of 66 Page 28 of 66 Page 28 of 66 Page 29 of 66 Page 29 of 66 Page 31 of 66 Page 31 of 66 Page 34 of 66 Page 34 of 66 Page 51 of 66 Page 51 of 66 Page 52 of 66 Page 52 of 66 Page 53 of 66 Page 53 of 66 Page 54 of 66 Page 54 of 66 Page 64 of 66 Page 64 of 66 Page 64A of 66 ATTACHMENT 8 Page 6 of 15 Page 6 of 15 Page 13 of 15 Page 13 of 15
T'
, INSTRUCTION SHEET
'Page 3
~ REMOVE (from TXX-88512)
INSERT (from TXX-88780)
ATTACHMENT 9 Page 2 of 42 Page 2 of 42 Page 3 of 42 Page 3 of 42-7 Page 4A of 42 Page 4B of 42 Page 5 of 42-Page 5 of 42 i
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Page 40 of 42 Page 40 of 42 i
ATTACHMENT 1G I
Page 5 of 54 Page 5 of 54 Page 10A of 54 Page 21 of 54 Page 21 of 54 Page 26 of 54 Page 26 of 54 Page 28 of 54 Page 28 of 54 Page 30 of 54 Page 30 of 54 Page 31 of 54 Page 31 of 54 Page 50 of 54 Page 50 of 54 Page 53 of 54 Page 53 of 54 Page 54 of 54 Page 54 of 54 ATTACHMENT 11 Page 6 of 52 Page 6 of 52 Page 8 of 52 Page 8 of 52 l
Page 9A of 52 Page 10 of 52 Page 98 of 52 Page 15 of 52 Page 10 of 52 Page 16 of 52 Page 15 of 52 l
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ATTACliMENT 13 6
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Page 8 of 8 Page 6 of 8 w - 1 Page 8 of 8 i
I f
l 1
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m.
m INSTRUCTION SHEET Page 4 REMOVE (from TXX-88512)
INSERT (from TXX-88780)
ATTACHffENT14
-Page 20 of 30 Page 20 of 30 ATTACHMENT 15 Page 2A of 20 Page 3 of 20 Tage 3 of 20 Page 8 of 20 Page 8 of 20 Page 16 of 20 Page 16 of 20 Page 17 of 20 Page 17 of 20 Page 18 of 20 Page 18 of 20 ATTACHMENT 16 Page 10 of 12 Page 10 of 12 ATTACHMENT 17 Page 4 of 41 Page 4 of 41 Page 7 of 41 Page 7 of 41 Page 8 of 41 Page 8 of 41 Page 8A of 41 Page 88 of 41 Page 18 of 41 Page 18 of 41 Page 19 of 41 Page 19 of 41 Page 20 of 41 Page 20 of 41 Page 23 of 41 Page 25 of 41 Page 26 of 41 Page 26 of 41 1
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ML20206G039

Text

SUBJECT:

Dearadation means a service-induced cracking,

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