ML20049H717
| ML20049H717 | |
| Person / Time | |
|---|---|
| Site: | Rancho Seco |
| Issue date: | 02/11/1982 |
| From: | SACRAMENTO MUNICIPAL UTILITY DISTRICT |
| To: | |
| Shared Package | |
| ML20049H714 | List: |
| References | |
| TAC-51696, NUDOCS 8203030479 | |
| Download: ML20049H717 (14) | |
Text
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RANCHO SEC0 UNIT 1 TECHNICAL SPECIFICATIONS Limiting Conditions for Operation 3.6.7 A mini-purge of the containment is permitted at the following reactor operating conditions:
N. Hot shutdown.
B.
Hot standby.
C.
Power operation.
3.6.7.1 Mini-purge shall be accomplished by lining up the following valves in the configuration described below.
Valve Number & Description Mini-purge Function Configuration SFV-53610 12 inch equalizing Mini-purge inlet Full open line valve SFV-53603 12 inch equalizing Mini-purge inlet Full open line valve SFV-53604 Normal purge Mini-purge outlet -< 50% open (<45 )'
outlet U
outlet
'-< 50% open (<45 )
SFV-53605 Normal purge Mini-purge outlet 3.6.7.2 Containment purging using the mini-purge system described in 3.6.7.1 shall be limited in duration to a maximum of 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> accumulated purging time per calendar month.
l B203030479 820211 PDR ADOCK 05000312 3-39a P
PDR Proposed Amendment No. 81
RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Limiting Conditions for Operation Bases The reactor coolant system conditions of cold shutdown assure that no steam will be formed and hence no pressure buildup in the containment if the reactor coolant system ruptures.
The selected shutdown conditions are based on the type of activities that are being carried out and will preclude criticality in any occurrence.
The Reactor Building is designed for an internal pressure of 59 psig and an external pressure of 2.0 psi greater than the internal pressure.
The design external pressure corresponds to the differential pressure that could be developed if the building is sealed with an internal temperature of 120 F with a barometric pressure of 29.0 inches of Hg and the building is subse-quently cooled to an internal temperature of 80 F with a concurrent rise in barometric pressure to 31.0 inches of Hg.
When containment integrity is established, the limits of 10 CFR 100 will not be exceeded should the maximum hypothetical accident occur.
The OPERABILITY of the containment isolation ensures that the containment atmosphere will be isolated from the outside environment in the event of a release of radioactive material to the contiament atmosphere by pressurization of the containment.
Containment isolation within the time limits specified ensures that the release of radioactive material to the environment will be.
i consistent with the assumptions used in the analyses for LOCA.
4 The use of a containment mini-purge at power provides a means to decrease containment tenperatures and airborne activity to increase the safety of personnel performing required surveillances inside the Reactor Building containment, and to comply with the objectives of ALARA.
Further, since the mini-purge may be used at power, there will be fewer occasions requiring a cold shutdown. This will reduce the number of thermal cycles that the vessel and RCS piping will experience.
REFERENCES FSAR, section 5 3-40a Proposed Amendment No. 81
Tabular torque data is calculated as follows:
45 Closure time = gge (5 sec.) = 2.5 sec.
To = Time from rupture.
To = 1.5 sec. + (45 - (valve opening)) 2.5 sec.
45 Ti = Time from initiation of valve movement.
Ti _ {45 - (valve opening)} 2.5 sec.
45~
Building pressure based on maximum building pressure ramp (FSAR Figure 14.4-5).
P = Building pressure.
33L P
0.5 7
1.0 13.5 1.5 19.5 2.0 24.0 2.5 28.5 3.0 32.0 3.5 35.0 4.0 37.5 4.5 40.0 5.0 42.5 5.5 44.5 6.0 46.5 Valve AP is derived from Allis-Chalmers Test A-C VER-0209 using appropriate values of valve configuration, upstream pressure and valve position.
C is derived from Allis-Chalmers Test A-C VER-0209 using appropriate values ofvalveconfiguration,upstreampressureandvalveposition.
Dynamic Torque = 166.4 C AP in accordance with A-C VER-0209 t
Bearing Torque = -76.93 AP in accordance with A-C VER-0209 Total Torque = (Dynamic Torque) + (Bearing Torque)
P.'.11.1 :G SLIP A!.a FROCF.Z1NG DiCWL, Foit s ECN A-8*2 7a2e 6
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FACE (S) WITM EXISTING E Q UI PM E N T.
IN DIC A T E WHAT NEW C O M PO N E N T(Sl ARE REQUIRED AND DR AWINGS AFFECTED.
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INDICATE WHAT NEW CO M PO N ENT($l ARE REQUtRED AND DR AWIN GS AFFECTED.
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DCN NUMBER IF NEW ORAWING BY SCC /GEC WRITE **NEW
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DCN ORAWING NO.
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ABE /55 O NUCLEAR C INSTRUMENTATION
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SACRAMENTO MUNICIPAL UTILITY DIST.RICT 4 OFFICE MEMORANDUM ll ~ *
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To:
D. G. Raasch DATE: flovember 19, 1980 e.
RJR 80-604 FRoM:
R. J. Rodriguez
' cuaJLcT:
Containment Bui.lding Integrity in th'e November 19, 1980 reply to Region V, the General' Manager stated that the final corrective action for 'the air-opera.ted containment purge valves would be taken during the 1981-refueling outage. This will require additional valves being procured and available by the time we shut down in May.
I request that Generation Engineering proceed with the final design and procurement so that installation can be completed, before our return to power in the late ' spring of 1981.
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CESCRIPTIOi:: /, design r.;o::if t:: tion tc ;roeide a failsafe cor.tition for valve SP!-53604 on less of diced :.urce:
pcwer, also for valves SPl-53503 and SFV-53610.
NCR ECi!
A-3?7,2 2.
SAFETY Ai!ALYSIS:
See attached Design Basis Report.
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i Licensing Engineer vate Manager Generation Engineering Date 3.
PRC RECOMPEf!DATION:
50.59(a)
Yes C No S 50.59(b)
Yes @ ?!o O DISPOSITION OF PRC:
a.
Unanimously recommends proposal 2 d.
Safety analysis inadequate O
b.
Send to MSRC for concurrence y
e.
MSRC review prior to implementing O c.
Recommends not to proceed O
f.
Test of system recuired g
$. b.0 &'bi PRC Chairman Date 4.
ATIALYSIS:
50.59(c)
Yes C Recommend to Proceed Yes @
No 0 No Refer to MSRC
~
50.59(b)
Yes g Test of system recuired @
No CLAdOx1 I - / P T 2_.
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Plant Sucerintencent Date 5.- MSRC FIftDINGS:
50.59(a)
Yes C No O 50.59(b)
Yes C No C DISPOSITI0il 0F MSRC:
i a.
Recommends proposal.
O c.
Recommends not to proceed O
._ b. Send to NRC for app'roval C d.
Safety analysis inadequate C MSRC Chairman Date _
6.
COMMISSI0il APPROVAL OBTAINED Date MSRC Chairman l
7.
RETEST COMPLETE AND ACCEPTABLE:
Test results approved Supervisor Enoineering and Ouality Control 8.
OVERALL REVIEM:
Plant modification change complete.
?!anacer :uclear Operaticns "a M__
9.
00CUME?iTATION COMPLETE:
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ENGINEERING ASSIGNMENT SHEET FOR ECN NO.A y 72
/
Date:
Design Basis Second Level Discipline Engineer Report Recuired Review Engineer Ctr' Cognizant Engineer 9)# ** '7 (2P~
8 M#
O Civil Engineer O
O Electric Design Engineer O
O I & C Design Engineer O
O Mechanical Design Engineer O
O Nuclear Design Engineer O
GEP"ffcensing Engineer (50.59s)
Bob Dieterich O
flW INSTRUCTIONS:
[,
1.
The Manager of Generation Engineering will assign the Cognizant Engineer and will indicate,in the square following the Cognizant Engineer's name, where a Design Basis Report is required.
2.
The Manager of Generation Engineering will also identify those ECN's that require a Safety Analysis Report by the Licensing Engineer, as well as the Second Level Review Engineer for the Safety Analysis.
3.
All other assignments required will be made by the Cognizant Engineer with the concurrence of the appro-priate Supervising Engineer. The Manager of Generation Engineering may make suggestions by marking certain boxes or identifying certain engineers for some or all of these activities, but these are only sug-gestions and should be treated accordingly. Under all conditions, the Cognizant Engineer must notify the personnel assigned and clear the assignment with the Supervising Engineer.
4.
The Design Basis Report may be prepared by the Cognizant Engineer or may be assigned to one or more design engineers on complex projects. Remember, each Design Basis Report requires a Second Level Review.
The Cognizant Engineer shall also assign these personnel, with the concurrence of the Supervising Engineer.
Wf&
D. G. Raasch Manager, Generation Engineering
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NCR N/A WORK REQUEST __J.04195 Discipline I&C Date Jan. 15,1932 I.
PURPOSE OF DESIGti C!IA!!GE:
The purpose of this design change is to implement the requirements of 10 CFR Part 50, Appendix A, Criterion 23 by providing valve closure on loss of direct current power.
This is further defined in the District's letter to the NRC, dated November 19, 1980.
" District engineers have designed a modification to the system which will provide a closure signal on loss of direct current power.
Upon completien of this modification, the valves in question will fully meet all the requirements of 10 CPR Part 50, Appendix A,
Criterion 23."
II.
DESIGN CRITERIA USED:
1.
Solenoid valves shall be Class IE and shall meet the requirements with respect to separation, seismic and Environmental Qualification.
2.
The following codes and standards are applicable to the solenoid valves purchase and installation.
Subject Sponsor Number IEEE 323 Quality Class 1E Equipment for Nuclear Power Generating S ta tions.
IEEE 344 Recommended practices for 2
Seismic Qualification of Class IE equipment for Nuclear Power Generating Stations.
III. CALCULATIONS AND DESIGN INFORMATION:
l.
Installation of the solenoid valves meet the requirements for Class IE safety g ra de.
Conduit runs and cables will remain as previously installed.
2.
Valves SFV 53503 and SFV 53504 require 3 solenoid valves in parallel to assure a 5 se cond closing of the time.
This is achieved by matchirg the Cy existing solenoid valves to the C
of the y
replacement solenoid valves, page 1 of 2
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3.
Valve SFV S3610 requires an exact replacement of the solenoid valve.
4.
The attached drawing 4-551 identifies the system configuration to meet the ImC requirement.
IV.
FAILURE MODE:
A credible failure resulting in a pipe break of any of the associated piping vill result in the loss of control air to the valve actuator and the associated containment isolation valve will fail closed.
A credible failure of any of the associated wiring (i.e.,
short circuit, open circuit) will result in failure of pouer to the valves which will then fail in the SFAS (safe) actuated state.
V.
COMMENTS:
The design meets the requirements of 10 CFR Part 50, Appendix A, Criterion 23 and is in compliance with the District's committment to the !IRC.
M-Date I-/3 '38-Design Engineer
.7yQ Review Engineer l.
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9332 SAORAMENTO f.tufilC; PAL UTILITY OtSTRICT C 6201 5 street. Boz I!320. Sauamesto, canternia 55alk (916) 452-3211 M. s -)
November 19, 1980
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Gen.,.
Mr. R. H. Engelken Director AEq w.,,?
Region V Office of Inspection & Enforcement IJ. S. Pluclear Regulatory Co mission 1990 tiorth California Soulevard Valnut Creek Plaza, Suite 202 Valnut Creek, Californla 94936 Re:
J. L. Crews to J. J. Mattimoe Letter Dated tJovember 7,1980 Operating License DPR-54 Docket tio. 50-31
Dear Mr. Engelken:
In reply to your letter from Mr. J. L. Crews reques ting an additional response to item C of the-tiotice of Violation dated September 26, 1980, we offer the following explanations and corrective actions to assure
, full. compliance witn flRC requirements.
Appendix A, i tem C of the September'26',1980 letter notes the following infraction:
10 CFR Part 50, Appendix A, Criterien 23 states that the protection sys tem shall be designed to fail into' a safe state or into a state demonstrated to be acceptable on e
some other defined basis if conditions such as disconnection
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of the. sys tem, loss of energy (e.g., electric power, instrument air), or postulated adverse environments (e.g. extreme heat or cold, fire, pressure, steam, water, and radiation) are experienced.
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In the FSAR Appendix 1A, Page IA-26 as part of a discussion on Criterion 23. the licensee states, " Safety features equipment can be ranually initiated by the operator at ar.y
' time even if power is lost to the actuation system."
Contrary to the above the reactor building purge inlet i
line valve SFV 53503, the reactor building purge outlet line valve SFV 53604, and the reactor building pressure equalizing line valve SFV 53210 will apparently fail open on loss of ' direct current power to the solenoid operators.
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Mr. R. H. Engelken N v;mb:r 19, 1980 SMUD Realv As stated above, the three valves do presently fail in the open
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position on a loss of direct current power to the centrolling solenoid valves.
Since the valves in cuestion are all outside containment valves wi th a fully redundant inside contain ent motor-operated valve, containment in tegri ty would not be violated upon loss of direct current power to the solenoid.
However, to further assure containment in tegri ty i t was decided to isolate the air to the three valves in ques tion.-enenever reactor building centainment integri ty is recai red.
Sasically, this " fails" the valves in their respective safety fea tures positions and subsequent loss of direct current power to the solenoic operaters will not alter the position of the valve.
The above corrective action has been *r plemented as an interim solution.
District engineers have designed a codificatioa to the system which will provide a closure signal on loss of direct current power.
Upon completion of this codificatien, the valves in question will fully meet all the require: ents of 10 CFR Part 50, Arpendix, A, Cri terion 23 I t is the District's intention to make these modifications during the 1981 refueling ou tage. -
Sincerely,
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Wm. C. Valbridge eneral Manager WCW:HH:jr p/
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