ML20235X309
| ML20235X309 | |
| Person / Time | |
|---|---|
| Site: | Comanche Peak  |
| Issue date: | 02/24/1989 |
| From: | William Cahill TEXAS UTILITIES ELECTRIC CO. (TU ELECTRIC) |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| TXX-89096, NUDOCS 8903130412 | |
| Download: ML20235X309 (10) | |
Text
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Loa # TXX-89096-File # 10014-
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C C-Ref.-# 10CFR50.36
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February 24, 1989-W.J.Cabm Esecutwe Vice President '
O. S. Nuclear Regulatory Commission Attn: Document Control Desk-Washington, D. C.
20555-
SUBJECT:
. COMANCHE PEAK STEAM ELECTRIC STATION (CPSES)
DOCKET N0. 50-445
. PROPOSED CHANGES T0.THE NRC PROOF & REVIEW TECHNICAL SPECIFICATIONS (SUPPLEMENT)
REF: William J. Cahill, Jr. letter to the NRC dated January 24, 1989 L
Gentlemen:
The referenced letter provided the NRC with TV Electric's proposed changed to the NRC Proof & Review Technical Specifications for CPSES Unit 1.
The following is a list of. attachments to this letter providing revisions to TV Electric's. proposed changes to address NRC staff concerns. -
T/S 3/4.4.5.2, " Operational Leakage," (Added footnote to clarify rearons for taking exception to valves.)
l Attachment e. -
T/S 3/4.6.1.2, " Containment Leakage," (Added: valves to be.
leak tested using water.)
Attachment.3 -
T/S 3/4.6.1.2 (Bases), " Containment Leakage," (Added valve leakage rates necessary to maintain a 30 day water seal.) -
T/S 3/4.8.1, "A. C. Sources," (Provided additional justification for the diesel generator overspeed trip i
setpoint.) -
T/S 5.3.2, " Control Rod Assemblies," (Include option for use of silver-indium-cadmium absorber material in control rods.)
If you require additional information concerning this revision, please contact this office.
Sincerely, 9
[
890313o412 890224
$DR ADOCK 0500 5
William J.
ahill, Jr
'JDS/grr.
q N[
c - Mr. R. D. Martin, Region IV g
Resident. Inspectors, CPSES (3) 400 North Olove Street LB 81 Dallas, Texas 7201
TXX-89096 Page 1 of,2 Q Q Q f p g y/g:.vy REACTOR C0OLANT SYSTEM OPERATIONAL LEAKAGE SURVEILLANCE REQUIREMENTS 4.4.5.2.1 Reactor Coolant Systas leakages, shall be demonstrated to be within each of the above limits by:
Monitoring the Reactor Coolant System Leakage Detection System a.
required by Specification 3.4.g.1 at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />; l Q.
.5 b.
Measurement of the CONTROLLED LEAKAGE to the reactor coolant puer seals when the Reactor Coolant Systes pressure is 2235 t 20 psig at least once per 31 days with the modulating valve fully open. The provisions of Specification 4.0.4 are not applicable for entry into MODE 3 or 4; g
Performance of a Reactor Coolant System water inventory balance at c.
least x:: ;:r ?* h: r g and Monitoring the Reactor Head Flange Leakaff Systas at least once per d.
24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
4.4.5.2.2 Each Reactor Coelant System Pressure Isolation Valveg shall be demonstrated OPERABLE by verifying leakage to be within ita limit:
a.
At least once per la months,.*
b.
Prior to entering.WDE 2 whenever the plant has been in COLD SHUTDOWN for 73h or more and if leakage testing has not been performedinthegrevious9 months,eaceptforvalves4701A,47018, 8702A,and87p.
j Prior to retu(rning'1ihe valve to service following maintenance, e
c.
repair 6r 7eplacement work en the valve, and d.
Fellowing check valve actuation due to flow through the valve.
As outlined in the ASE Code,Section XI, paragraph IW-3427(b).
e.
The provisions of Specification 4.0.4 are not applicable for entry into M DE 3 or 4.
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TXX-89096 Page 2 of 2-l, J-e INSERT FOR PAGE 3/4 4-15 This exception allowed since these valves have control room position indication, inadvertent opening interlocks and a system high pressure alarm.
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Mttachment 2
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TXX-89096~'
-Page 1 of 2 CONTAIMENT SYSTEMS SUR EILLANCE REQUIREMENTS (Continued) t b.
.If any periodic Type A test fails to meet either 0.75 L, or 0.75 L '
i t
the test schedule for subsequent Type A tests shall be reviewed and approved by the Commission.
If two consecutive Type A tests fail to meet either 0.75 L, or 0.75 L, a Type A test shall be performed at t
least every la months until two consecutive Type A tests meet either at which time the above test schedule may be 0.75 L, or 0.75 Lt resumed; c.
The accuracy of each Type A test shall be verified by a supplemental test which:
1)
Confirms the accuracy of the test by verifying that the supple-mental test result, L,, is in accordance with the appropriate following equation:
lL - (L, + L,) l 1 0.25 1.a or l L ~ blte * 'o) l $ 0.25 Lg g
c where L,or Ltm is the measured Type A test leakage and L,,
is the superimposed leak; 2)
Has a duration sufficient to establish accurately the change in leakage rate between the Type A test and the supplemental test; and
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3)
Requires that the rate at which gas is injected into the contain-ment or bled from the conta6nment during the supplemental test
'e ~
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and 1.25 L
- is between 0.75 L, and 1.25 L,; or 0.75 Lt t
-3.
d.
Type 8 and C tests shall be conducted with gas at a pressure not' less than P,, 48.3 psig, at intervals no greater than 24 months except for tests involving:
1)
Air locks,.
7 2)
Containment ventilation _ isolation valves with resilient material
- seals,
/ Is sp=Wisa In Qs,J.P%kt= V.6.1.2.h a J
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Safety Injection Valves 4-0000A
,.cd Containment Spray Valves % g = f..t.1 "."^Tp" W h 46.
3) a,s 4a 0, 17!,
, 10-l'2,..cd 10-446.
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Air locks shall be tested and demonstrated OPERA 8LE by the require-
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ments of Specification 4.6.1.3;
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Containment ventilation isolation valves with resilient material seals shall be tested and demonstrated OPERA 8LE by the requirements ofSpecification4.6.1.7. gor 4.6.1.7.j,asapplicable;(t TA5ERI' g.
Safety Injection Valves 4-8800A,1 :^ 2, ;.741 ^^"" C 1j L i :.:tr:t:d ^."!?'"'I b; ;;rf:r ::: Of : :^^- l: " : :::rr;- ::', a tt'!: pr;;;;r'::d *- a pressure not less ttwn i ND>sig, at intervals no greater than g,
/.iik,,53.33 h..
Containment Spray Valves 1HV-4776,# 1HV-477bICT-142, ~, 1CT-145 shall be leak tested with water, at a pressure not less'than F, C'".3bsig, at intervals no greater than 24 months; and 1.l P., (3.3 3 The provisions of Specification 4.0.2 are not applicable'L.s / /
i.
COMANCHE PEAK - UNIT 1 3/4 6-3
Attachment' 2 TXX-89096 Page 2 of'2 INSERT FOR PAGE 3/4 6-3 1-8802A, 1-8802B, 1-8809A, 1-88098, 1-8818A, 1-8818B, 1-8818C, 1-88180, 1-8819A, 1-88198, 1-8819C, 1-8819D, 1-8835, 1-8840, 1-8841A, 1-8841B, i
1-8905A, 1-8905B, 1-8905C, and 1-89050 shall be leak tested with water at v
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1 a
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proof f REVIEW 3/4.6 CONTA! WENT SYSTEMS BASES 1
3/4.6.1 PRIMARY CONTAIMENT 3/4.6.1.1 CONTAINMENT INTEGRITY Primary CONTAI MENT INTEGRITY ensures that the release of radioactive materials from the containment atmosphere will be restricted to those leakage paths and associated leak rates assumed in the safety analyses. This restriction, in conjunction with the leakage rate limitation, will limit the EXCLUSION AREA B0UNDARY radiation doses to within the dose guideline values of 10 CFR 100 during accident conditions.
3/4.6.1.2 CONTA!WENT LEAXAGE The limitations on containment leakage rates ensure that the total l
containment leakage volume will not exceed the value assumed in the safety analyses at the peak accident pressure, P,.
As an added conservatism, the measured overall integrated leakage rata is further limited to less than or equal to 0.75 L, or 0.75 L, as applicable, during perfomance of the periodJc7 t
test to account for possible degradation of the containment leakage barriers between leakage tests.
[ e M6ERT /
A The surveillance testing for measuring leakage rates is consistaqt with j ~j the requirements of 10 CFR 50 Appendix J.
3/4.6.1.3 CONTAIMElff AIR LOCKS The limitations on closure and leak rate for the containment air locks are required to meet the restrictions on CONTAllgENT INTEGRITY and containment leak rate. Surveillance tasting of the air lock seals provides assurance that the overall air lock leakage will not become excessive due to seal damage during the, intervals betwee3 air lock leakage tests.
3/4.6.1.4 INTEm mL PRES $URE The 11eitattene en containment internal pressure ensure that: (1) the containment stmeture-is prevented from exceeding its design negative pressure differential of 5 paid with respect to the outside ataesphere, and (2) the contai m nt peak pressure does not exceed the design pressure of 50 psig.
duringgLOCA.
The maximum peak pressure aspected to be obtained from a LDCA event is M#
U".1: ;;te. TM !!:f t Of L5 ;;%.: psig, which is less than design pressure
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COMANCHE PEAK - UNIT 1 8 3/4 6-1
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.TXX-89096 Page 2 of 2 l
INSERT FOR PAGE B 3/4 6-1 For specific system configurations, credit is taken for a 30 day water seal that will be maintained to prevent containment atmosphere leakage through the penetrations to the environment.
The following is a list of the containment isolation valves that meet this system configuration and the Maximum Allowed Leakage Rate (MALR) required to maintain the water seal for 30 days.
Valve No, MALR Valve No.
MALR Valve No.
MALR 1-8802A (later) 1-8819A (later) 1-8905A (later) 1-8802B (later) 1-8819B (later) 1-8905B (later) 1-8809A (later) 1-8819C (later) 1-8905C (later) 1-88098 (later) 1-8819D (later) 1-8905D (later) 1-8818A (later) 1-8835 (later)
ICT-142 (later) 1-88188 (later) 1-8840 (later)
ICT-145 (later) 1-8818C (later) 1-8841A (later) 1HV-4776 (later) 1-88180 (later) 1-8841B (later) 1HV-4777 (later) i
i TXX-89096 Page-1 of 1 JUSTIFICATION FOR FREQUENCY LIMIT FOR LOAD REJECTION l
TECHNICAL SPECIFICATION 3/4.8.1 The diesel generator overspeed trip setting is 115% of rated speed as recommended by. the vendor for nuclear emergency diesel. generators. The basis for 115% overspeed setpoint is to meet the IEEE 387 requirement for full load rejection without tripping on overspeed. The frequency limit of'60 i 6.75 Hz for the load rejection surveillance test is based on 75% of the difference
' between nominal (60' Hz) and the overspeed trip (69 Hz) which is less than 15%
above nominal in accordance with the Standard Technical Specifications (NUREG-
'0452, Rev. 4).
- Attachment 5 qwp y4gy,u..
N9'IN1bNFEATURES O
c.
, Nominal thickness of concrete walls = 4.5 feet.
- ; 'v
- d.
Nominal thickness of concrete roof = 2.5 feet, e.
Nominal thickness of base mat = 12.0 feet.
f.
Nominal thickness of steel liner wall = 3/8 inch.
(Dome = 1/2 inch, Base Mat = 1/4 inch),, and Netfreevolume=(2,985,000fcubic. feet.
g.
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.0ESIGN PRESSURE AND TEMPERATURE 5.2.2 The containment building is designed and shall be maintained for a maximum internal pressure of 50 psig and a temperature of 280*F.
5.3 REACTOR CORE FUEL ASSEMBLIES 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 to e initial core loading and shall have a maximum enrichment not to exceed weight percer.t U-235.;
g-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 stee1 t king.
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5.4-REACTOR COOLANT SYSTEM m,3,,,,, r. 4 c.
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I DESIGN PRESSURE AND TEMPERATURE g m e m ihy G rd-5.4.1 The Reactor Coolan ystem is designed and shall be maintained; InaccordancewiththeCoderhquirementsspecifiedinSection5.I a.
of the FSAR, with allowance for normal degradation pursuant to the applicable Surveillance Requirements, b.
For a pressure of 2,485 psig, and c.
For a temperature of 650*F, except for the pressurizer which is 680*F..
O COMANCHE PEAK - UNIT 1 5-4 i
m_
'TXX-89096*
Page 2 of 2 e
JUSTIFICATION FOR CHANGE TO TECHNICAL SPECIFICATION 5.3.2 CPSES FSAR allows the use of hafnium or silver-indium-cadmium (Ag-In-Cd) absorber control rods.
It is possible.that CPSES will utilize Ag-In-Cd and/or -
hafnium control rods. Hafnium and Ag-In-Cd control rods both exihibit acceptable neutron absorber and dynamic mechanical response characteristics.
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