ML17299A561
| ML17299A561 | |
| Person / Time | |
|---|---|
| Site: | Palo Verde  |
| Issue date: | 08/30/1985 |
| From: | Van Brunt E ARIZONA PUBLIC SERVICE CO. (FORMERLY ARIZONA NUCLEAR |
| To: | Knighton G Office of Nuclear Reactor Regulation |
| References | |
| RTR-REGGD-01.052, RTR-REGGD-1.052 ANPP-33312-EEVB, NUDOCS 8509050041 | |
| Download: ML17299A561 (87) | |
Text
REGULATO INFORMATION DISTRIBUTION STEM (RIDS)
ACCESSION NOR:8509050041 DOC ~ DATE: 85/08/30 NOTARIZED: YES DOCKET FACIL:STN-50-529 Palo Verde Nuclear Station< Unit 2< Arizona Publi 05000529 STN"50-530 Palo Verde Nuclear StationE Unit BYNAME 3E Arizona Publi 05000530 AUTH AUTHOR AFFILIATION VAN BRUNTPE.E, Arizona Nuclear Power Project (formerly Arizona Public Serv
~
RECIP ~ NAME RECIPIENT AFFILIATION KNIGHTONE GiN ~ 'Licensing Branch 3 SUOJ/CT; Forwards FSAR changesrincluding deviation from Rev 1 to Reg Guide 1.52 8 rev to 10CFR50 73 re nonroutine
~
reportability requirements. Changes will be included in next FSAR update.
DISTRIBUTION CODE: BOOID COPIES RECEIVED:LTR $ ENCL Q SIZE':
TITLE: Licensing Submittal: PSAR/FSAR Amdts 8, Related Correspondence NOTES:Standardized plant. 05000529 Standardized plant. 05000530 RECIPIENT COPIES RECIPIENT COPIES ID COD /NAME LTTR ENCL ID CODE/NAME" LTTR ENCL NRR/DL/ DL 1 0 NRR LB3 BC 1 0 NRR LO3 A 1 0 LICITRAEE 01 1 1 INTERNAL: ACRS 6 6 ADM/LFMO 1 0 ELD/HDS3 0 IE FILE 1 1 IE/DEPER/EPB 36 1 1 IE/DQAVT/QAB21 1 1 NRR ROE<M,L 1 1 NRR/DE/AEAB 1 0 NRR/DE/CEB 11 1 1 NRR/DE/EHEB 1 1 NRR/DE/EQB 13 2 2 NRR/DE/GB 28 2 2" NRR/DE/MEB 18 1 1 NRR/DE/MTEB 17 1 1 NRR/DE/SAB 20 1 1 NRR/DE/SGEB 25 1 1 NRR/DHFS/HFEBQO 1 1 NRR/DHFS/LQB 32 1 NRR/DHFS'/PSRB 1 1 NRR/DL/SSPB 1 0, NRR/DSI/AEB 26 1 1 NRR/DS I/ASS 1 1 NRR/DSI/CPB 10 1 1 NRR/DS I/CSB 09 1 1 NRR/DS I/ICSB 16 1 1 NRR/DSI/METB 12 1 1 NRR/DSI/PSB 19 1 1 NRR/DSI/RAB 22 1 1 NRR/DSI/RSB 23 1 1 E 04 1 1.
RGN5 3 3 R i)I /MI8 1 0 EXTERNAL: 24X 1 1 BNL(AMDTS ONLY) 1 1 DMB/DSS (AMDTS) 1 1 LPDR 03 1 1 NRC PDR 02 1 1 NSIC 05 1 1 PNl GRUELER 1 1 TOTAL NUMBER OF COPIES REQUIRED: LTTR 52 ENCL
e P II J
n
Arizona Nuclear Povver Project P.o. BOX 52034 ~ PHOENIX, ARIZONA85072-2034 Director of Nuclear Reactor Regulation August 30, 1985 Attention: Mr. George M. Knighton, Chief ANPP-33312-EEVB/JKO Licensing Branch 3 Division of Licensing U. S. Nuclear Regulatory Commission Mashington, D.C. 20555
Subject:
Palo Verde Nuclear Generating Station (PVNGS)
Units 2 and 3 Changes to PVNGS FSAR on Various Subjects Docket Nos. STN-50-529/530 File: 85-056-026'.l.01.10
Dear Mr. Knighton:
Attached for your review on PVNGS Units 2 and 3 are FSAR changes to various sections.
These changes are editorial in nature. They are being submitted for your review.
Listed below are the subjects addressed by the FSAR changes:
(1) Deviation from R.G. 1.52 Rev. 1 to be consistent with Tech Specs.
(2) Reg. Guide 1.16 has been superceded by a revision to 10CFR 50.73 with regard to non-routine reportability requirements.
(3) Conform with as-built system configuration for seismic, monitors.
(4) Correctly identify plant vent radiati'on monitors as tag numbers XJ-SQN-RU-143 and XJ-SQN-RU-144.
(5) Revise Figure 3.6-4 to correctly cross reference break number 2206 to Figure 3.6-2.
(6) Amend FSAR references to Chapter 16 t'o redirect reviewer to the Tech Specs.
For PVNGS Unit 1, safety reviews and evaluations have been completed for implementation of these changes in accordance with the requirements of 10CFR 50.59.
The safety reviews and evaluations have determined that there are no unreviewed safety questions involved with the changes. These changes will be included in the next FSAR update.
gnl 850905004 PDR j 850830 ADOCK 05000529,'
A .: ., PDR ~.*I
,G~ N. Knighton
, Changes to PVNGS FSAR on ious Subjects ANPP- 33312 Page 2 1'f you have any questions concerning these changes, please contact William Quinn of my staff.
Very truly yours, E. E. Van Brunt, Jr.
Executive Vice President Project Director EEVB/JKO/slh Attachment cc: E. A. Licitra M. Ley R. P. Zimmerman A. C. Gehr
'Il STATE OF ARIZONA )
) ss.
COUNTY OF MARICOPA)
I, Edwin E. Van Brunt, Jr., represent that I am Executive Vice President, Arizona Nuclear Power Project, that the foregoing document has been signed by me on behalf of Arizona Public Service Company with full authority to do so, that I have read such document and know its contents, and that to the best of my knowledge and belief, the statements made therein are true.
\
QC U Edwin E. Van Brunt, Jr.
Sworn to before me this89 day of , 1985.
Notary Public My Commission Expires:
My Commission Expires April 6, 1S07
t G.~ M. Knighton Changes to FSAR on Vario Subjects ANPP- 33312 Page 3 bcc: D. B. Karner J. G. Haynes R. M. Butler M. E. Ide E. C. Sterling A. C. Rogers M. F. Quinn T. F. Quan LCTS Coordinator S. R. Frost K. M. Gross J. M. Allen J. R. Bynum
- 0. J. Zeringue J. Orlowski L. G. Papworth J. D. Houchen S. Shapiro C. F. Ferguson SARCNs: 2005, 2007, 2015, 2022, 2038, 2040
PVNGS FSAR CONFORMANCE TO NRC REGULATORY GUIDES Position C.3.o
.Air straightening devices are installed only if tests indicate that uniform air flow distribution is not achieved.
Position C.4.b Vacuum breakers would be of minimal assistance in opening doors to units during fan operation. The use of vacuum breakers creates potential leakage paths.
Units which operate at higher pressure than external pressures; i.e'., push-through units, do not require vacuum breakers.
K. Position C.4.c Accessibility for ease of maintenance is provided by removing opposing filters in opposite directions. The .
standard suggested distance of 3 feet plus length of component for removal of filters is met.
Position C.4.d Piping associated with manifolding could result in plate-out of components of the sampled gas stream, ggsaa~ 6 leading to erroneous test results. The test, probes are located in readily accessible locations with a minimum of piping, but are not manifolded.
Position C.5.d The activated carbon adsorber section will be leak 12 tested in accordance with position C.5.d of R.G. 1.52 rev 2 March 1978 and using ANSI N510-1980 in place of ANSI N510-1975.
Amendme~" 12 1.8-368 February 1984
Insert A to FSAR Page 1.8 - 368 M. Position C.4.e Each atmosphere cleanup train will be operated for 15 minutes per month.
There is not expected to be any moisture buildup on the absorbers and HEPA filters due to the low humidity at PVNGS . This is in agreement with the Palo Uerde Unit 1 Technical Specifications which require the systems to be operated for at least 15 minutes every 31 days.
I PPg. cQ.UN 7
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PVNGS FSAR CONFORMANCE TO NRC REGULATORY GUIDES REGULATORY GUIDE 1.13: Fuel Storage Facility Design Bases (Revision 0, March 10, 1971)
RESPONSE
The position of Regulatory Guide 1.13 is accepted. Compliance is described in sections: 9.1.2, Spent Fuel Storage; 9.1.3, Spent Fuel Pool Cooling and Cleanup System; 3.5, Missile Pro-tection; 3.8.4, Other Seismic Category I Structures; and 9.4, Air Conditioning, Heating, Cooling, and Ventilation Systems.
REGULATORY GUIDE 1.14: Reactor Coolant. Pump Flywheel Integrity (Revision 0, October 27, 1971)
RESPONSE
Refer to CESSAR Section 1.8.
REGUIATORY GUIDE 1.15: Testing of Reinforcing Bars for Category I Concrete Structures (Revision 1, December 28, 1972)
RESPONSE
The position of Regulatory Guide 1.15 is accepted (refer to section 3.8).
REGULATORY GUIDE 1.16: Reporting of Operating Information (Revision 1, October 1973)
RESPONSE
The position of Regulatory Guide 1.16 is accepted, exceptae-Foa sEc:(ma Q R> Nciv'eaonius EEpogvr, uJhluk
/g ga~ygecy'eo~ toWX.W >2 ~~~'< 73 f<iC, ~Id&~'V~E 4+< r'<~"
February 1984 1.8-7 Amendment 12
...Two'eismic triggers are provided to start the SMA recording----
system. One is located adjacent to the SMA in the containment tendon gallery and the other on the containment operating floor at elevation 140 feet. Both triggers are sensitive to
...vertical..and...horizontal. motions-."-
A magnetic tape recording and.'playback unit is provided for multiple channel recording and playback of the signals from
'he strong motion accelerometers. The data recordings
" include,hM additional recording channel *whi'ch"contains a signal. Internal to the ta e recorder are t 'ers "'iming which continuously monitor the accelero utputs and Tc4sagT Q activate the recording system if the remote seismic trigger(s) should fail.
The recording and playback system is housed in a panel furnished for,these. instruments and devices necessary for system testing, annunciating, calibration, and control. ,This panel is located in the control room.
3.7.4.2.2 Peak Recording Accelerograph The sensor unit contains three -accelerographs mounted in a mutually orthogonal array. The PRA is mounted directly on Class 1 pipe in the auxiliary building at elevation 78 feet and has one axis coincident with the principle pipe axis.
It is " self-powered unit.
The PRA is located as necessary to verify the continued availability of Seismic Category I systems and equipment.
Data from the PRA must be manually retrieved following an earthquake and are used in the detailed investigations for particular structures, systems, and equipment.
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Amendment 6 3.7- '0 October 1981
PVNGS FSAR APPENDIX 9A b) Samples are available by placing the tank in a recircu-lation lineup and drawing a sample of the pump dis-charge. This is described in section 11.2.2.2.4.1.
Descrete sample capabilities are listed in table 9.3-3 (sheet 10 of 12).
c) The concentrate monitor tanks are kept in a continuous recirculation mode while they contain radioactive concentrate. 'amples 'are taken from an analysis point off the pump discharge. Sample capabilities are listed in table 9.3-3 (sheet 10 of 12).
d) The fuelstorage area currently is designed to be grab sampled for iodine using a movable airborne monitor (refer to table 11.5-1). Continuous iodine collection using a charcoal cartridge will be pro-vided for the fuel building.
The radvaste area and evaporator vent are designed to be grab sampled from the grab sam le connection on radiation monitor XJ-SQN-RU-14 ra.lian'so~
rior to release, the plant vent<monxtors (XJ-S U-13 ) continuously and isokinetically collects an iodine sample using a charcoal cartridge. Additionally, areas in the radwaste building may be grab sampled for iodine using a movable airborne monitor.
(NRC Question 460.6) (9 4 and 11 3)
Provide a table comparing the design features and radio-activity removal capability of each normal ventilation filter system to each position detailed in Regulatory Guide 1.140, Rev. 1 (October 1979), "Design, Testing and Maintenance Criteria for Normal Ventilation Exhaust System Air Filtration and Adsorption Units of Light-Water-Cooled Nuclear Power Qg5,ALT A May 1981 9A-25 Amendment 4
8 LOWDOWN LINES TO SG NO. 1 ISG E439 AND 053)
~ IPE BREAK LOCATION BREAK NVMOE4 IVHIP AESIAAINT TARGET 0 2 6 4 MINAL f NOS 241) NONE NONE 2411 NONE NONE 2418 NONE NONE 2421 NONE NONE 2422 NONE NONF 0 MAX ST4ESS LOCATIONS 2416 NONE NONE 2424 NONE CONTAINMfNT PENETAATION 2424 241) 2427 JET IMPINGEMENT ON SLOWDOWN LINES (SG.E4)39 AND 053)
ORICFNOF JET 84EAK NVMBER TARCET SG-EH)63 SG.E-012 TOWETLAVVP RECIAC
~2416 12 ACS I)02 IFIGVRE36 10) 3.4%6 fROM WET LAVVP ACS I I ) I IfIGV RE 1 6-10) 12.13 AECIAC 4CS l)12 IF IGVAE36 'IO) 3.4+ g BSLI O,I I, ACS l)11IFIGVAE 16-10) 89,10.11,13.14.15 BI 1%
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B.t'-Z LE CE NOI' PIPE BREAK LOCATION O PIPE WHIP 4 ES14AINT JET IMPINCEMENT TARGET PAIU VPFIIB Nue(PAF GPAPFS) in')81(an YSAI(
S/G No. 1 BLO'4T)OWN LINES (SC-E-039 nND O531 uenT)ON or rosT(II+TED BRrJ)Krolt)Ts.
JET IN@It)G)JIENT TnRGrTS ANo rlrE HNIr RESTRn(NTS rigvre 3.5-8 Dqcnn)~r 1983 nnen<lnont lo
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CONFORMANCE 10 NRC REGULATORY GUIDES emergency procedure is implemented. This ensures proper operator response independent of event diagnosis. This approach is consistent with CEN-152,
'CE Emergency Procedure Guidelines.
I Actions identified as subsequent operator action are addressed as a recovery procedure, implemented after event diagnosis. This approach is consistent with CEN-152, CE Emergency Procedure Guidelines.
The specific procedure format and content has been identified in the Emergency Procedure Generation Package and submitted to the NRC for review. This is consistent with NUREG 0899.
The implementation of the positions of this Regulatory Guide are described in chapters 13~and 17> and +be Tacan'irei.
Spec,4caWone.
February 1985 1.8-21B Amendment 14
PVNGS. FSAR CONFORMANCE TOOHRC
'-REGULATORY GUIDES tl hih> j.ihYi"~ci'i'lU Dr". 8 REGULATORY GUIDE 1.34: Control of Electroslag Meld Properties (Revision 6, Di cember 28, 1972)
RESPONSE. "
to CESSAR
'efer Section 1.8.
REGULATORY GUIDE 1.35: Inservice Inspection of Ungrouted Tendons in Prestressed Concrete Containment "Stiuctures (Revision 1,
RESPONSE
--June 1974)'"-'e~+
P/Q ( ( ~$ '~ +Qc4Y)accLL +PQciP~$ 6 S~
Inservice surveilla e requirements are discussed in surveillance program complies w'th Regulatory Guide 1.35, Revision 1, except, for the following:
Position C.S.b Position CD 5.b is replaced with Position C.4 of Regu-latory Guide 1.35, Draft Revision 3. In addition, as the elongations at installation and during inspection may vary more than 5% due to creep, shrinkage, thermal effects, and friction, a recording of the elongation throughout the history of the inspection program will be maintained so that trends can be identified. This data will be used in evaluating f any unusual differences in elongation during installation and inspection.
Position C.5.c Position C.5.c is replaced with Position C.2 of Regu-latory Guide 1.35, Draft Revision 3.
1.8-22
CONFORMANCE VO NRC REGULATORY GUIDES
-- governor oil'ooler. This system is not part of .the diesel generator unit.
- ..B.. Position C.2.e.7 Tests to verify correction of a problem will be con-ducted after the affected diesel is declared "ready for service." The diesel and associated systems may be operated as necessary to perform troubleshooting and verify correction. of specific problems, prior to such declaration,, without these operations counting as a test, for the purposes of complying with this Regulatory Guide.
A Qp ec Flccc&b As~
a REGULATORY GUIDE 1.111: Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Effluents in Routine Releases from Light-Water-Cooled Reactors (Revision 1, July 1977)
RESPONSE
Information contained in Regulatory Guide 1.111 is utilized as discussed in section 2.3 REGULATORY GUIDE 1.112: 'Calculation of Releases of Radioactive Materials in Gaseous and Liquid Efflu-ents from Light-Water-Cooled Power Reactors (Revision O-R, May 1977)
RESPONSE
Information contained in Regulatory Guide 1.112 is utilized as discussed in section 11.3.
1.8-64
PVNGS FSAR CONFORMANCE TO Ng REGULATORY GUIDES lb ~ ll ~ ~ ~ 'L ar ' ~
this item will be satisfied provided that such work is performed onnly w'ith approved procedures and that the activities and the'esults are documented. Evidence of design change approval shall be required prior to placing the affected item in service.
D. Section 5:
For the purposes of functional tests addressed by this standard', APS" defines completed sy'stems as any system, or portion or component thereof; on which construction is suffic'iently'complete to allow the required testing,
'I and on which further or adjacent construction will not render the 'results of such testing invalid or indeterminate.
E. Item 5.1.g:
'Tr'aceability as used in this item is considered to be the same as discussed in section 5.2.13.3 of ANSI N18.7.
REGULATORY GUIDE 1.117: Tornado Design Classification (Revision 1, April 1978)
RESPONSE
The position of Regulatory Guide 1.117 is accepted to the extent described in sections 3.3, 3.5 and .9.2.5.4.
REGULATORY GUIDE 1.118: Periodic Testing of Electric Power and
'Protection Systems (Revision 1, November 1977)
RESPONSE
The in sections 7.1g ..
position of Regulatory 00 Al& gec4gicctL 5pegigcck065 Guide 1.118 and.
is accepted as described Q.3> awk Sector S/Q.B Yiay 19 81 3..8-65B Amendment 4
PVNGS 'SAR PROTECTION AGA'INST:DYNAMIC EFFECTS ASSOCIATED WITH, THE POSTULATED RUPTURE OF PIPING
- 2. ,-The system or~portion of.a system sustaining the leakage..crack operates during normal plant operas o ~n.l defined in dl ~LR,'0 modes 1, 2, and 3 1 L 4'%p
- 3. The failed line is greater than 1-inch in diameter.
B. Where a postulated leakage crack occurs in one train of a Seismic Category I, dual-purpose, moderate-energy piping system, single active component failures are not assumed in the other train (Refer to Branch Technical Position APCSB 3-1, B.3.b.3). The postulated leakage crack must not adversely affect active components of both trains.
C. Through-wall leakage cracks are not postulated in portions of ASME Code,Section III, Class 2 moderate energy fluid syst: em piping passing through the contain-ment penetrations and extending to the first outside isolation valves if they meet the requirements of NE-1120 of ASME Code Section III and the combined
, stresses, as calculated by Equations (9) and (10),
Paragraph NC-3652 of the ASME Code,Section III, do not exceed 0.4 (1.2 Sh + SA).
D. In portions of ASME Code,Section III, .Class 2 and 3 piping and non-nuclear piping located within, or out-side, and adjacent to protective structures containing safety-related systems or components, through-wall leakage cracks are postulated where combined stresses, as defined previously, exceed 0.4 (1.2 Sh.+ S ) except as exempted in sections 3.6.2.1.3.1.C and 3.6.2.1.3.1.E.
The cracks are postulated to occur individually at locations that result in the maximum effects from fluid spraying and flooding, and the consequent hazards or environmental conditions developed.
3.6-40
Ih
~ ~ IV ~ 0 ~
APPENDIX 4A the test is conducted routinely and not just during tne initial low-power'physic's'-"'te'sti'ng.'""'ESPONSE:
PVNGS will conduct the CEA symmetry test des-cribed in CESSAR Section 14.2.12.4 during the low-power physics testing conducted during initial startup and following each subsequent core reload.
K *' . ("",Q" "" (4.2)
Please describe the details of the visual surveillance which will be conducted on PVNGS discharged fuel assemblies. This surveillance program should be adequate to identify gross problems of structural integrity, fuel rod failure, rod bowing,
\
spacer grid strap damage, insufficient fuel rod shoulder gap spacing; or crud deposition on a limited number of fuel assem-blies that are discharged at each refueling. If abnormalities are found during the examination, the licensee should agree to write an LER; .otherwise, no reporting or licensing action is necessary.
RESPONSE: PVNGS shall visually inspect a limited number of randomly selected (about 10 to 15) discharged fuel assemblies during or following each refueling. The visual inspection shall be conducted with underwater viewing equipment (will include inspection of the four sides of each inspected fuel assembly) and is intended to detect gross problems of structural integrity, gross fuel rod failure, gross bowing, spacer grid strap damage, insufficient, fuel rod shoulder gap spacing, or crud deposition. Underwater viewing equip- k ment is separately provided on the refueling machine and in the spent fuel pool. Reporting of abnormalities will be in accordance with Sections l' Amendment 6 October 1981
"'VNGS F SAR INTEGRITY C4'EACTOR COOLANT PRESSURE BOUNDARY "6.2.-2..7 Material= S ecification .
'l Refer to CESSAR Section 5.4.13. In addition. for material specifi'cations related to the secondary system overpressuriza-tion protection refer to section 10.3.2
~ ~
5.2.2.8 Process Instrumentation
. Refer to CESSAR, Section 5.2.2.7. In addition. refer to fig-ures 5.1-1 and 10.3-1 for the instrumentation related to pri:mary and secondary system overpressurization protection,
.respectively.
5 '.2.9 S stem Reliabilit Refer to CESSAR Section 5.2.2.8. Also, refer to sections 5.1.5 and 10.3.2 for a discussion of secondary system overpressure reliability.
. protection 5.2.2.10 Testin and Ins ection Testing and inspection of primary and secondary valves are governed by ASME Section XI. Testing and inspection of the CLH d secondary safety valves is discussed in sections 3.9 <14.2 and +6 .:144 Sec+on >/'f.'7 l oC +h~ Te&nica) Spec~-CicaHoAS 5.2.2.11 Over ressure Protection Durin Low Tem erature Conditions Refer to CESSAR Section 5.2.2.10 except as specifically modified by the following changes:
o For PVNGS-specific Pressure-Temperature curves refer to the PVNGS Technical Specifications.
o CESSAR Figure 5.2-1 is superseded by figure 5.2-7.
~ CESSAR Figure 5.2-2 is superseded by figure 5.2-8.
~ CESSAR Subsection 5 ' ' '0.2.3 pressure relief valve setpoint is 46/ pslg ~
February 1985 5.2-23. Amendment 14
P.VNGS FSAR INTEGRITY OF RESISTOR COOLANT PRESSURE BOUNDARY
~ a a E" ~, ~ ~ ~ . ~ %PI L ll 5.2.3.4.2 Control of Welding 5.2.3.4.2.1 Avoidance of Hot Crackin A... Components in C-E Scope of Supply Refer to CESSAR Section 5.2.3.4.2.1-A.
B. 'omponents Not in C-E Scope of Supply In order .to preclude microfissuring'n austenitic stainless steel, PVNGS design is consistent with the recommendations'of Regulatory Guide 1.31 except as noted in section 1.8.
5.2.4 INSERVICE INSPECTION AND TESTING OF REACTOR COOLANT PRESSURE BOUNDARY Details of the inservice inspection program are included in section 6.6>and . . Accessibility of inspection areas is
~
discussed in CESS Section 5.2.4.1.
g, y,p/q.q,Z o5 Ae. Wed c l Spec &Mons.
5.2.4.1 DELETED 5.2.5 REACTOR COOLANT PRESSURE BOUNDARY LEAKAGE DETECTION SYSTEMS Means for the detection of leakage from the reactor coolant pressure boundary are provided to alert operators to the exis-tence of leakage above acceptable limits, which may indicate an unsafe condition for the facility. The leakage detection systems are sufficiently diverse and sensitive to meet the criteria of Regulatory Guide 1.45 for leaks from identified and unidentified sources.
5.2.5.1 Leaka e Detection Methods 5 2.5.1.1
~ Unidentified Leakage The four methods employed to detect unidentified leakage are presented in the following sections.
Amendment 12 5.2-136 February 1984
PVNGS FSAR l INTEGRITY OF RECTOR COOLANT PRES SURE BOUNDA'RY n
5.'2.5.2.3.2 'Sum 'Level 'Measurin S stem. The initiation of an additional or abnormal leak in the .containment results in an increase in the flow rate to the sump. The additional sump flow initiates an excessive leakage flow alarm in the control room.
Upon actuation of the excessive leakage alarm, the operator follows these steps:
A. Records sump level measurements at equal time i'ntervals.
il B. Determines the sump level increase during these intervals.
C. Determines the leakage from the measured sump level increases.
~
- Section >jt 't.< o~
5.2.5.3 Limits for Reactor Coolant Leaka e Refer to CESSAR Section 5.2.5.3. Also, refer to+the technical specifications for RCS leakage.
5.2.5.4 Maximum Allowable Total Leaka e Refer to CESSAR Section 5.2.5.4. Also, refer to~the technical specification limits for identified leakage.'.2.5.5 Differentiation Between Identified and Unidentified Leaks Refer to CESSAR Section 5.2.5.5. The systems used to detect unidentified leakage from the RCS to the containment are described in section 5.2.5.1.1.
5.2.5.6 Sensitivit and 0 erabilit Tests Refer to CESSAR Section 5.2.5.6. Additionally, a description of tests to demonstrate the operability of the leakage detec-tion systems are provided in section +6.~~ 5jd "i + oC 'ita'-
Tachvi>Cal 'Speci Sicahon5.
February 1984 5. 2-143 Amendment 12
';~" i R EACTOR- E S SE E
- 5. 3. 1.7: Reactor Vessel. Fasteners Ref ei to'(.ESSAR 'Section 5.3;1.7.
'I
'I Fracture toughness and tensile test data for reactor vessel
~ ~
closure head bolting are presented for Palo Verde Unit-1 in table 5.3-9, for Unit 2 in tables 5.3-12 and 5.3-13, and for Unit 3 in tables 5.3-21 and 5.3-22.
Fracture toughness and tensile test data for all other fasten-ers used in the reactor coolant pressure boundary (RCPB) for Palo Verde Unit-1 are presented in tables 5.3-10 and 5.3-11, for Unit 2,in. tables 5.3-14 to 5.3- 20, and Unit 3 in tables 5.3-23 to 5.3-29.
5.3.2 PRESSURE TEMPERATURE IiIMlTS Refer to CESSAR Section 5.3.2.
pressure- temperature limits are discussed 3l't 78 A,e. Wec4>'ca( Spec! Q~b~nS.
5.3.3 REAC."r~)R Vl SSEL tNTEQR1TY R< f <! r to CESSAR Section 5. 3. 3.
5.3.3.1 I)<'. i.gn Refer to CESSAH Section 5.3.3.1.
5.3.3.2 Materials of Construction Hefer to CESSAR Section 5.3.3.2.
5.3.3.3 Fabrication Methods Refer to CESSAH Section 5.3.3.3.
Refer to CESSAR Section 5.3.3.4.
Amendment 14 5.3-46 February 1985
~:-
REACTOR VESSEL 5.3.3.5 Shi ment and Installation Refer to CESSAR Section 5.3.3.5. For a discussion of compli-ance with Regulatory Guides 1.37 and 1.39 during installation, refer to section 1.8.
r 5.3.3.6 0 eratin Conditions Refer to sections 3.9 and 4.4 for information on design transients and operating conditions, respectively.
5.3.3.7 Inservice Surveillance Refer to CESSAR Section 5.3.3.7.
For a discussion of the inservice inspection program soe unb sections 6.2.~446.6. and ~~" 5ec&on +/ IA'.9 crW &<
Tec4vica.lj, 'Spec~ Qca+c 05, 5.
3.4 REFERENCES
"C-E Procedure for Design. Fabrication, Installation and Inspection of Surveillance Specimen Holder Assemblies," Combustion Engineering Topical Report, CFNPD-155P, Approved August 11, 1975.
Amendment 14 5.3-78 I'ebruary 1985
PVNGS FSAR COMPONENT AND SUBSYSTEM DESIGN Operational/Controls=.: ==- =:-.-
- 1. - The SCS'components shall be powered such that the
~
-'operational and control requirements of Section 5.4.7.1.3 (A) are met.
- 2. The SCS shall meet the operation and control requirements of Section 5.4.7.1.2.
'. Inspection and Testing
- 1. All SCS ASME,Section III components shall be arranged to provide adequate clearances to permit.
inservice inspection.
- 2. Manually operated valves which contain reactor coolant or other potentially radioactive liquids during normal plant operations shall be provided with handwheel extensions and shielding, to allow periodic actuation.
- 3. SCS components which contain reactor coolant or other potentially radioactive liquids during normal plant operations, and which require access
" for periodic pressure tests and nondestructive examination, shall be capable of being flushed prior to testing. The low pressure safety injec-tion pumps shall provide the driving head for flushing.
System components not designed to ASME, Section III, should be located such that the access for periodic visual inspection for leakage, structural distress, and corrosion is possible.
- 5. System and component arrangement shall allow ade-quate clearances for performance of inspections identified in Technical 'SPec'~~~~~+~
February 1984 5.4-65 Amendment 12 l12
,.PVNGS FSAR 3/q q,g ~/+he, We hn'c <
$ pp g, Q~~Q on5 g(>c rYI Lflg APPENDIX 5A
- 3. Include the v'i additional
' i ~ ~ I .
requirements listed in NUREG-0212 regarding eddy current testing in Section 4.4.5.2.b.3;
~ ~
- 4. Change the wording
~ I ~
in Section 4.4.5.3.b to be consistent with the corresponding in NUREG-0212;
- 5. Include additional requirement listed in NUREG-0212 regarding preservice inspection in Section 4.4.5.4.a.9;
- 6. Include additional requirement listed in NUREG-0212 with regard to the reporting requirements in Sections 4.4.5.5.a; 4.4.5.5:b,"4:4;5:5.c.
RESPONSE: The response is given in~amended sec-tion 16.3/4.4.5.,
(5;2)
Specify the edition and addenda to which all reactor coolant pressure boundary components were fabricated.
RESPONSE: The response is provided in amended table 5.2-0.
To demonstrate compliance with the beltline material test requirements of Paragraph III.C.2 of Appendix. G, 10 CFR Part 50.
a) Provide a schematic of the reactor vessel, showing all welds, plates and/or forgings in the beltline. Welds should be identified by shop control number, weld pro-cedure qualification number, the heat of filler.metal, and type and batch of flux. Provide the chemical composition for these welds (particularly Cu, P, and S content). Identify material specification, type and grade of all base metal.
October 1981 5A-3 Amendment 6
PVNGS FSAR APPEND I 5A certified by qualified supervisory personnel. Records of the certification of personnel are maintained and available for review at C-E's Chattanooga facility.
Individuals performing inservice fzacture toughness tests shall be qualified by training and experience and shall have demonstrated competency to perform the tests in accordance with written procedures and ASME Code,Section III, Sub-article NB-2300,. Fracture Toughness Requirements for Mate-rials..'he zecommendations for qualification of nuclear power plant inspection, examination, and testing personnel that are included in ANSI N45.2.6 - 1973 are generally acceptable and provide an adequate basis for complying with Paragraph III.B.4 of Appendix G, 10 CFR Part 50.
F,-~.,e ~/~ 3.<-z .<
Q ( Q""
Provide pressure temperature limit, curves for the reactor pressure vessel.
are RESPONSE: Pressure temperature curves~4 '&provided in>
the Technical Specifications, I /
lt ** " " ( Q"""
Provide the following data on the surveillance materials:
a) Origin of heat affected zone and base materials (heat number, plate identification number, and chemical composition),
b) Origin of weld metal (weld wire type, heat of filler metal, production welding process, plate material used to make weld specimens, chemical composition of deposited weld metal),
c) The lead factor of each surveillance capsule with respect vessel inner wall.
RESPONSE: The response is given in amended section 5.3.1.6 and table 5.3-4.
Amendment 7 SA-8 December 1981
PVNGS'SAR'ONTAINMENT SYSTEMS B. The c'ontainment spray system consists of two redundant and independent trains each of which is capable of providing 100% of the required heat removal capability and l00% of the required iodine removal capability.
C. . The heat. removal capacity of the CSS is discussed in section 6.2.1.
D. The containment heat removal system is designed to remain operable in..the containment accident environ-ment as discussed.. in section 3.11.
E. '. The CSS is'designed such that single failure of any
'.active component will not degrade system abilities as shown in the failure modes and effects analyses of CESSAR Appendix 6A, Table 4.5 (for further discussion of system actuation, see section 7.3).
F. 'he 'enti;re CSS is designed to Seismic Category I
'equirements.'ystem components as appropriate are designed to meet ASNE Code Section III, Class 2 requirements.
The CSS is protected against dynamic effects associated with postulated rupture of piping as discussed in sec-t-'on 3.6. 3/5k.2, qg Ae Technical Cpeclkcabons The CSS is desi ed to permit. the periodic inspections and tests described in section< . and CESSAR Appendix 6A, Sections 7.10, 8.2 and 9.0.
The CSS is designed to add hydrazine to the spray water to reduce fission product 'odine concentration in the containment atmosphere as discussed in se'ction 6.5.
The CSS is sized based on the long-term heat rejection function of the system. The shutdown cooling heat exchangers used for rejecting heat from the contain-ment are sized by the shutdown cooling function discussed in section 5.4.7.
6.2.2-8
- >'-'.PANGS.::F.SAR CONTAlNMENT SYSTEMS Tests Bhye~e&ified",that. the:hfdf'Ogden'=oxygen rec'ombination i:s not a catalytic"surface, effect..associated with the heaters, but occurs'ue to the increased temperature of the process gases.
As the phenomenon is.not a catalytic effect, saturation of the unit is not predicted to occur. Results of testing a prototype electric hydrogen recombiner and production unit test results are given in reference 1. There is no difference between the hydrogen recombiner units to be installed in PVNGS and the unit for wh'ich the tests were conducted.
6'.2.5.2:;2.2:' dro en'Monitorin Subs stem. The hydrogen monitoring"subsyst'em for each unit consists of two completely redundant'trains. Each train consists of a hydrogen sensor, an electronic subassembly and local and remote readout/alarms.
The electronic subassemblies for trains A and B are housed separately in cabinets located in the auxiliary building.
A bottled ni:trogen and *hydrogen supply is used to calibrate the sensors at tnose intervals specified in section h Hydrogen measurement is accomplished by using a thermal con-ductivity cell and a catalytic reactor. The sample gas first flows through the sample section of the cell, then passes through the catalytic converter wher hydrogen in the sample is catalytically combined with free oxygen to form water. vapor, then passes through the reference section of the cell. The nydrogen content is indicated by the difference in thermal con-ductivity between the sample and reference sides of the cell.
Oxygen, in an amount sufficient to combine hydrogen at the highest range of the analyzer, is added to the sample gas, prior to passing through the sample section of the cell. The range and accuracy of the hydrogen analyzer is given in table 7.5-1.
A single failure analysis is given in table 6.2.5-2.
( q. ~p Qchg'iccA Spec'<PincN+~> ~
p/q g Amendment 9 6.2 '-10 August 1982
,.C.ONTA $ NMENT, S YR'EIUS reactor coolant system (short term), and containment sump water (long term) would provide the heat sources which establish and maintain upward natural convection flows within the containment. The water level in the 4
containment post-LOCA has been calculated to be approximately 10 feet 6 inches above the floor and has a nominal temperature of 300F.
6.2.5.4 Test and Inspections The analytical and .test program for the hydrogen recombiners.
includes proof-of-principle tests and full-scale prototype tests on a production recombiner. The tests were completed and the results of these tests were submitted to the NRC in reference 1.
In the design of the equipment actually installed at PVNGS, all recombiner components can be inspected and are accessible for maintenance during normal plant operation.
Periodic testing of the containment hydrogen control system is described in Initial testing is described in section 14.2.~ @ ~ >/~ ~ q q g ye g; ( g eo;Poohong, t
6.2.5.5 Instrumentation Re uirements 6.2.5.5.1 Hydrogen Recombiner Subsystem A manual control station is provided for each train for start-ing and stopping the unit. The controller maintains the correct power input to bring the recombiner above the threshold temper-ature for the record>ination process. The controller setting can be adjusted to accommodate variations in containment tem-perature and pressure in the post-LOCA environment. The system is designed to conform to the applicable portions of IEEE-279 and is powered from a Class IE source. No automatic initiat-ing signals or alarms are provided. The hydrogen recombiners 6.2.5-23
/
rP,VNGS FSAR EMERGENCY CORE COOLING SYSTEM
~ s e I r ~
- 3. Two independent vital instrument power sources
'shall be provided for the SIS instrumentation-.--
See A.S above.
Thermal Limitations
~ W ~ I Not Applicable Monitoring
- 1. Provisions shall be made .for the detection, containment, and isolation of the maximum expected
--- leakage'rom'a moderate energy pipe rupture, as
'i.'scussed in" C.l above.
- 2. 'rocess instrumentation shall be available to the operator in the control room to assist in assess-ing post-LOCA conditions. The type of instrument, parameter measured, instrument range and accuracy ar'e listed in Table 6.3-3. I Operational Controls Not Applicable Inspection and Testing
- 1. Inspection and testing requirements for the SIS are contained in Section ~Mand shall be
,COmplied With. (>/q ~ g ~ yeghioef gpedgCakalS
- 2. Prior to initial plant startup, SIS flow tests shall be performed. An adequate supply of water and the necessary test connections't the containment sump shall be provided.
Chemistry/Sampling
- 1. The Sampling System shall provide a means of obtaining remote liquid samples from the SIS for chemical and radiochemical laboratory analysis.
6.3-5
PVNGS FSAR HABITABILITYSYSTEMS
'1'v'o .lndegen }en'(; v~ a,I ) ns v.ru]penj 4f>>J~.
f c)p( ~ ~ ..
Design and testing'f filtration systems is p consistent. with the recommendations of NRC Regulatory Guide 1.52, Design, Testing, and Maintenance Criteria for Atmosphere Cleanup System
<I Air Filtration and Adsorption Units of Light Water-Cooled
~ 4 Nuclear Power Plants, except as discussed in section 1.8.
Inservice testing of the control zoom essential HVAC system is specified in section ~~
conducted in accordance with the surveillance requirements 3/H.'7.'7 of +4e. Teohnicnl CpeclgcskonS Portable equipment such as air samplers, personnel dosimeters, and other radiation analysis equipment applicable to control room habitability is tested and inspected periodically as noted in section 12.5.
6.4.6 INSTRUMENTATION REQUIREMENT The following indications are displayed in the control room:
fan status, damper positions, room temperatures, and outside air intake radioactivity. Alarms indicate open access'doors ls after transfer to the emergency mode, low fan differential pressure, and outside air intake airborne radioactivity great'er than 10 -6 pCi/cm 3 (Xe-133). (Refer also to sections 7.3
~
and 7,.5.)
Instrumentation required for actuation of the control room essential HVAC system is discussed in section 6.4.2.2.2 and in section 7.3. System control logic diagzam is shown in figure 7.3-6.
Details of the radiation monitors used to initiate CREFAS are given in sections 7.3 and 11.5. Information including detector locations, type of radiation detected, detector type, range, and sensitivity are given in table 11.5-1.
The instrumentation is designed as Seismic Category I. A des-cription of initiating circuits logic interlocks and periodic testing requirements and redundancy of instrumentation relating to control room habitability appears in section 7.3.
March 1982 6.4-27 Amendment 8
PVNGS'SAR / P FISSI i PRODUCT REMOV AND CONTROL SYSTEMS test 'of"a'epresentative'sample'of the impregnated activated=-.
charcoal is performed to verify iodine removal efficiencies.
~ ~
Design andtesting of,ESF. filtration systems is consistent
'I with the. recommendations of NRC Regulatory Guide 1.52, Design, Testing and Maintenance Criteria for Atmosphere Cleanup System Air Filtration and Adsorption Units of Lightwater-Cooled Nuclear Power Plants, as discussed in section 1.8.
Preoperational testing is performed on systems in accordance with'he test descriptions in section 14.2.
.q s~., ~ < <<
6.5.1.4.2 Inservice Testing Inservice testing of the ESF filtration sy tems is conducted in accordance with the surveillance requirements>of the technical specifications,
~~
6.5.1.5 Instrumentation Requirements Controls and instrumentation for the control room and for the fuel building systems are discussed in section 7.3. Each system is designed to function automatically upon receipt of an ESF actuation system signal. Fans can also be controlled from the control room.
The status of the essential ventilation equipment is displayed in the control room during both normal and accident operations'ection 1.8 addresses the extent to which the recommendations of NRC Regulatory Guide 1.52 are followed with respect to instrumentation.
6.5.1.6 Materials The materials of construction used in or'n the filter systems are given in sections 6.4.2.2 and 9.4.5.2. Each of the materials is compatible with the normal and accident environ-ments postulated in the control room and the fuel building.
6.5-6
~ ~
F-I.SSI~ PRODUCT- REIIOVAL ~
AND CONTROL SYSTEMS.
-.. E. " .
Pressure differentials .between compartments are 1'nsignificant.
-:F, LOCA mass/energy sources are identified as. being below
" the operating .floor, or approximately the lower 25% of
~
.; ...- the containment volume.
G. Sprayed and unsprayed volumes each are homogeneous in terms of concentrations and distribution of mass/energy.
sources.
The PVNGS 'design:utilizes .'230 Spraco 17071417 (15.2 gal/min) spray nozzles in each train of the primary spray headers. It also uses 80 Spraco 17651308 (3 gal/min) nozzles in each train f8
'of the auxiliary spray headers.
6.5.2.4 Tests and Inspection 6.5.2.5 f 'd'thy' Preoperational testing is performed on the system in accordance with the, test. description in section 14.2. Periodic testing sedlon 3/Q.@.2, o0 0+e. Technica( 5pecigcakons.
Instrumentation Requirements The iodine removal system is provided with instrumentation and controls to allow the operator to monitor the status of the system. All instrumentation, with the exception of pressure instrumentation, receives emergency onsite power from separate, redundant, and train-aligned power supplies.
Level indication is provided locally and in the control room to monitor spray chemical storage tank (SCST) availability. A low level alarm will denote loss of the stored hydrazine
/
solution. A low-low level signal will stop the spray chemical addition pumps (SCAP) and will close the SCST isolation valves. Level switches are also provided to close the IRS isolation valves at the low-low SCST level setpoint.
March 1982 6.5-11 Amendment 8
E PVNGS FSAR FISSION PRODUCT REMOVAL AND CONTROL SYSTEMS (RA)7.6.4 For mechanical independence, see (RA)7.3, (RA)7.4, and (RA)7.5.
(RA)7.7 THERMAL LIMITATIONS Each CSS train is provided with an independent environmental control system, such that the safety-related equipment in each train operates within the
- environmental design limits specified in
-:"" CESSAR Section"3.11.
(RA)7.8 MONITORING Provisions are made for detection, con-tainment, and isolation of the maximum expected leakage from a moderate energy p.'pe rupture in each train.
Redundant pressure, temperature, and flow instrumentation is available to the operator in the control room to assist in assessing post-LOCA conditions. The types of instrument, parameter measured, instrument range and accuracy are listed in CESSAR Section 6.2.
(RA)7 ' OPERATIONAL AND CONTROLS See (RA)7.1.
(RA)7.10 INSPECTION AND TESTING For inspection and testing, see 2/q
- z. C+I, .T.J '
~i g ca'h czA 8.
6.5-50
///s ~ ~ / /
INSTRUCT'KNTAT/ON~AND;CONTROLS 7.1.2.7 Conformance to IEEE 338
/ ~ M II Refer to'ESSAR Section 7.1.2.7.
In addition, in conformance to IEEE 338-1971, response time testing for all PPS channels and equipment is performed during pre-operational testing and each refueling interval. Sec-tion ~describes testing freguency.
7.1.2.8 Conformance to IEEE 344 Refer to CESSAR Section 7.1.2.8.
Conformance to IEEE 344-1975, for non-C-E scope, is discussed in section 3.10.
7.1.2.9 Conformance to IEEE 379 as Au ented b Re lator Guide 1.53 Refer .to CESSAR Section .1.2.9.
In 'addition, the entre Pps s discussed iri section 7.1.1.1 and essential safetgreglted supporting systeras listed in section 7.1.1.4 comp y wi the requirements of IEEE 379-1972 as augmented by Regulatory Guide 1.53. The single failure criterion is discussed in section 7.3.2.
7.1.2.10 Conformance to IEEE 384 as Au ented b Re lator Guide 1.75 Refer to CESSAR Section 7.1.2.10.
1 In addition, compliance to General Design Criterion 17, IEEE 384-1974, and Regulatory Guide 1.75 is described in section 8.3.
7.1.2.11 Conformance to IEEE 387 Conformance to IEEE 387-1972 is discussed in section 8.1.
7.1-4
PVNGS FSAR INSTRUMENTATION AND CONTROLS 7.1.2.30 Conformance to Re lator Guide 1.100 Conformance to Regulatory Guide 1.100 is presented in sec-tion 3.10.
7.1.2.31 Conformance to Re lator Guide 1.105 Conformance to Regulatory Guide 1.105 is presented in sec-tion 1.8. Z/g.~ ~g +4e. Mechanic~( %Pe~'~
7.1.2.32 Conformance to Re ator Guide 1.118 Conformance to Regulatory Guide 1.118 is given in sec-tion 1.8 and implemented in section ~-.Bj~ Specific test capabilities within the reactor protective system and the engineered safety features systems are described in sec-tions 7.1.2.7, 7.2.1, and 7.3.1.1.
7.1.3 CESSAR INTERFACES The following NSSS general interface requirements are repeated from CESSAR Section 7.1.3.
7.1.3. 1 Power Vital instrument power requirements for the safety-related systems are discussed in Section 8.3.1.
7.1.3.2 Protection from Natural Phenomena Refer to Section 3.1.2. CESSAR Design Scope Class 1E equipment shall be located within the plant so as to ensure the various natural phenomena specified in GDC 2 which are applicable to the Applicant's site will not result in degradation of that equipment below the level required to allow it to perform required protective action assuming a single failure.
- 7. 1-8
PVNGS 'SAR INSTRUMENTATION'ND CONTR(k'S" 7.1 3.10 Ins ection and'Testin g, +or/w.> oP The PPS, including sensors, shal be capable of being periodi-cally tested in accordance with>the Technical Specifications Those portions which could adversely affect reactor operations shall be capable of being tested when the reactor is shut down. All other safety-related instrumentation shall be capable of being tested during. normal operation.
7.1.3.11 Chemist Samolin The, components of the safety-related equipment shall be located so as not to exceed the chemistry limits specified in
~
section 3.11.
7.1.3.12 Materials Not applicable to the safety-related instrument and controls equipment.
7.1.3.13 S stem Component Arran ement.
Safety-related components shall be located so as to conform to the separation, independence, and other criteria'pecified in this section. The safety-related components shall be located to provide access for maintenance, testing and operation as required.
Analog and digital signals provided to the safety-related components shall not share the same multiconductor cable, unless specifically called for or approved by Combustion Engineering.
7.1.3.14 Radiolo ical Naste Radiological waste discharge lines or components shall not be routed or located next to protection system electronic compo-nents in a manner that will result in exceeding the radiation limits specified in section 3.11.
M 7.1-11
PVNGS FSAR 7.2 REACTOR PROTECTIVE SYSTEM Z/q.g-o0 Ae Ted,n A 7.
2.1 DESCRIPTION
" gpect QCck~
Refer to CESSAR Section 7.2.1.1, except for the following additional discussions:
A. The required response time testing discus d in CESSAR Section 7.2.1.1.9.8 is addressed in section A,M~+.
The methods, equipment and test frequency are also r I dS'p B. ~
Refer..to..section..8.3 .for a discussion of how the
,.PVNGS. design meets, the CESSAR interface requirements for vital instrument power supplies.
C. The Core Protection Calculator (CPC) System and CEA Calculators described in CESSAR Section 7.2.1.1.2.5 also provide their outputs and a number of their inputs as inputs to the Plant Monitoring System (PMS) by means of fiber-optic data links. The values transmitted represent the raw sensor count data.
These values are converted to engineering units within the PMS. The data is transmitted to the PMS at least once every ten minutes.
II The CPC/CEAC Data Link data processing programs within the PMS perform cross-channel comparisons for each input signal and generate an alarm whenever the difference between any single channel's value and the average value of all four channels is greater than a constant. On operator demand, a report is printed to show the results of the latest cross-channel comparison. The CPC and CEAC parameters are not used by any other program in the plant computer.
Refer to CESSAR Section 7.2.1.2.
Instrument location layout drawings are presented in fig-ures 7.2-1, 7.2-2, 7.2-3, and 7.2-4.
February 1984 7.2-1 Amendment 12
'PVNGS F SAR REACTOR PROTECTIVE SYSTEM interacts'on," the 'PVNG'S .de'si'gn-.'also provides a mono-directional
.data.. link.,from .the, Core Protection Calculator (CPC) System to the Plant Monitoring, System by means of fiber-optic data links.
These 'data links are identical to the hardware utilized at each CEA Calculator output (See CESSAR Section 7.2.1.1.2.2). The non-conducting fiber-optic cable used ensures that no electrical failure at the Plant Monitoring System will affect the Core Protection Calculators or the CEA Calculators.
7.2.2.3.3 Testing Criteria E
Refer to CESSAR .Section 7.2..2.3.3. In .addition, for .the...
organization .for testing and documentation, refer to chap-ter 13. Minimum frequencies for checks, calibration and peri-odic testing are given in section-.&~~ 3/6,5.l oW +he Tee'en~~~
+pic Qca+cAS.
7.2.2.4 Failure Modes and Effects Anal sis Refer to CESSAR Section 7.2.2.4.
7.2.3 REACTOR PROTECTIVE SYSTEM (RPS) INTERFACE REQUIREIKNTS The following interface requirements are repeated from CESSAR Section 7.2.3.
The interface requirements discussed below are specific to the RPS. General interface requirements are discussed in sec-tion 7.1.3.
7.2.3.1 Power Vital instrument power interface requirements are discussed in section 8.3.1.
February 1984 7.2-2A Amendment 12
'PVNGS 'SAR ENGINEERED SAFETY FEATURE SYSTEMS
. "Table 7;3-1;"-
>ONE-OUT,.OP~TWO ESFAS. BYPASSES,,
..Title .. Function Initiated By Removed By Trip Channel
- .; "Bypass:~a~
Disables any Manually by Same switch
- ; .given trip controlled channel access switch
- a. Interlocks allowed only one channel for any type trip to be bypassed at one time
..--is acceptably..low. during. maintenance. bypass periods. The "bypass's manually initiated and manually removed. An electri-cal interlock allows only one channel for any one type trip to be bypassed at one time. Bypasses are annunciated visually and audibly to the operator.
In some cases, bypass of more than one parameter within a channel- may be required in the eve~t of an equipment failure.
/'k 3.2.
Weekni~h 5 p ec~ ~ ~4~~.
7.3.1.1.3.2 Operatin B passes. For two-out-of-four operat-ing bypass capability refer to CESSAR Section 7.3.1.1.3.2.
For the'one-out-of-two logic there are no operating bypasses.
- 7. 3. l. 1. 4 Interlocks For two-out-of-four interlocks refer to CESSAR Sec-tion 7. 3. l. 1. 4. The one-out-of-two ESFAS interlocks prevent the operator from bypassing more than one trip channel for a type trip at a time. Different type trips may be bypassed simultaneously, either in the same channel or in different channels.
7.3-6
~
~ 'I ENGINEERED SAFETY 4 r, FEATURE SYSTEMS B. The equipment can perform as required.
.....-,C =...-The.-interactions -of-protective -actions, control actions, and. the environmental changes that. cause, or are caused by, the design basis events do not prevent the mitigation of the consequences of the event.
D. The system cannot be made inoperable by. the inadvertent actions of. operating and maintenance personnel.
In addition, the design is not encumbered with additional components or channels without re'asonable assurance that such additions are beneficial.
7.3.1.1.7 Sequencing There is no sequencing for any ESF equipment other than that necessary'or ESF'bus loading. The automatic load sequencer is discussed in section 8.3.1.1.3.
7.3.1.1.8 Testing For two-out-of-foui'testing capabilities see CESSAR Sec-tion 7.3.1.1,8. In addition, provisions are made to permit periodic testing of the one-out-of-two ESFAS. These tests cover the trip actions from sensor input through the pro-tection system and the actuation devices. The system test does not interfere with the protective function of the system.
The testing system meets the criteria of IEEE Standard 338-1971 and of Regulatory Guide 1.22.
Since actuation of the ESF systems controlled by the one-out-of-two ESFAS does not disturb normal plant operating conditions, the one-out-of-two ESFAS is tested by complete actuation as described below. Frequency of'ccomplishing the tests is 1' '
~3/t.S.2 C A i Qpggi+Cp+< O ~ <g +4~ Tgcknicch 5Pec 7.3-9
PVNGS FSAR ENGINEERED SAFETY FEATURE SYSTEMS r
the ESFAS signal:--Mai'n'tenance 'a'nd calibration of the bypassed
'hannel can be accomplished in a short time interval. Probabil-ity"of fail'ure o'f the remainin'g channel is acceptably low during such mainte'nance periods.'
4.12 "Operating Bypasses There are no operating bypasses.
4.15 Multiple Setpoints There are no multiple setpoints.
~ \
4.21 System Repair Identification of a defective channel will be accomplished by observation of system status lights or by testing as described in section 7.3.1.1.8. Replacement or repair of components in the actuation logic is accomplished with the affected channel bypassed. The affected trip function then operates in a single active channel trip logic.
7.3.2.3.3 Testing Criteria IEEE Standard 338-1971 and Regulatory, Guide 1.22 provide guidance for development of procedures, equipment, and docu-mentation of periodic testing. The basis for the scope and means of testing are described in this section. Test intervals and their bases are included in section . . . The organi-A zation for testing and for documentation is described in chapter 13. Since operation of the ESF system is not expected, the systems are periodically tested to verify operability.
Complete channels can be individually tested without violating the single failure criterion and without inhibiting the oper-ation of the systems. The system can be checked from the sensor signal through the actuation devices during reactor operation since ESF system operation does not damage equipment or disturb reactor operation. Thus, testing completely simulates valid actuation.
gpec,R~Honh ~
7.3-55
PVNGS FSAR ENGINEERED SAFETY I ~ E ~ T FEATURE SYSTEMS'I Minimum frequencies for checks, calibration, and periodic testing of. the ESFAS instrumentati.on and control ar'e given in "
t section... S/+ 3.2 o< +he Te'chniM Spq/g+Q+l'oA5.
The use of individual trip and ground detection lights, in conjunction with those provided at the "supply bus, ensure that possible grounds or shorts to another source of voltage will be detected.
'he response time from an input signal to protect system trip bistables through the opening of the actuation relays is veri-fied by measurement-during plant startup testing. Sensor responses are measured during factory acceptance tests. Sec-tion 7.3.1.1.8.3 provides additional information on response time testing.
7.3.2.4 Failure Modes and Effects Anal sis Refer to CESSAR Table 7.2-5. The failure modes and effects analysis for the additional ESF systems is given in tables 7.3-14 through 7.3-18.
7.3.3 CESSAR ESFAS INTERFACE REQUIREMENTS The. following interface requirements are repeated from CESSAR Section 7.3.3:
The interface requirements discussed below are specific to the ESFAS.
General requirements are discussed in Section 7.1.3. Those items specific to the RPS are discussed in Section 7.2.3.
7.3.3.1 Power Refer to Section 8.3.
7.3 ' ' Protection from Natural Phenomena Refer to Sections 3.1.2 and 7.1.3.2 ~
7.3-56
'l y PANGS FSAR SYSTEMS REQUIRED...FOR
,SAFE SHUTDOWN 7; 4. 1. 2 .-. Desi Basis "Information Reefer* to CESSAR'ection'7;4.1.2 for design bases of the systems in the CESSAR scope. For systems not in the CESSAR scope, refer to the design bases discussion in the appropriate section. of this chapter. '
7.4.1.3 Final S stem Drawin s For final system drawings in the CESSAR scope, refer to CESSAR Section 7.4.1.3. In addition, section 1.7 includes a list of applicable electrical and instrumentation drawings and piping and instrumentation, diagrams which have been provided.
Furthermore, equipment location layout drawings are included in section 1.2.
7.4.2 ANALYSIS Refer to CESSAR Section 7.4.2.
7.4.2.1 'onformance to IEEE 279-1971 In addition to the analysis provided in CESSAR Section 7.4.2, certain IEEE 279-1971 criteria, not in the CESSAR scope, are addressed below (heading numbers correspond to the section numbers in IEEE 279-1971).
4.5 CHANNEL INTEGRITY See section 14.2 for preoperational test, procedures.
4.10 CAPABILITY FOR TEST AND CALIBRATION see section /13.5 and~&for periodic testing procedures. geo~on n/Q g o$
ep.c.~ F'a %on~
~ 'Techni ~L 7.4.2.2 Conformance to IEEE 308-1971 Refer to CESSAR Section 7.4.2.2.
Amendment 13 7.4-12 August 1984
..PVNGS FSAR ALL OTHER INSTRUMENTATION SYSTEMS REQUIRED FOR SAFETY to determine if cooling water is available to each pump and to take appropriate action in less than 30 minutes to protect the reactor coolant pump affected.
The instrumentation is provided in compliance with the require-ments of IEEE Standard 279-1971.
7.6.2.1.3.2 Safet In ection Tank Pressure Monitorin . Moni-toring the SIT pressure with two visual status alarms for each channel on low SIT pressure provides information to the operator to determine the unavailability of the SITs to perform their core flooding function in the event, of a LOCA. The instrumen-tation is provided in compliance with the requirements of IEEE 279-1971.
7.6.2.1.3.3 Auxiliar Buildin ESF Puma Room Level Monitorin Monitoring each ESF pump room level with one visual status alarm for each room provides sufficient information and 30 minutes of time for the operator to take appropriate action to prevent equipment flooding at a leakage rate of 50 gal/min. The instru-mentation is provided in compliance with the requirements of IEEE-279-1971, except for the redundancy requirements.
7.6.2.2 Anal sis of E ui ment Desi n Criteria 7.6.2.2.1 Shutdown Cooling System Suction Line Valve Interlocks Refer to CESSAR Section 7.6.2.2.1. In addition, for periodic testing requirements, see for access procedures for setpoint adjustment , calibration, and test points, see section 13.5.
g/q p ~g ~e T(chnicnl GpeciAakons)
I" 7.6.2.3 . Fire Protection Instrumentation and Detection S stem An analysis of the fire protection system is discussed in section 9.5.1.
February 1985 7.6-5 Amendment 14
Oh) PVNGS:,FSAR .
PgiN7 ONSITE PONER SYSTEM&
'It tt r B. 1nterrupted auxiliary feedwater flow to the steam generator(s) is fully reestablished within 23 C-14 seconds. The deviation from the CESSAR requirement
- of "15 seconds is acceptable to Combustion Engineering 0 +he.
's discussed in sectio 1.9.2.4.10.
~
regni cJ 6p?cc Qcahof 5
)
for testing requirements.
- 3) During testing if an SIAS or AFAS occurs while the diesel generator is paralleled to the preferred power supply with the 3f control switch in the REMOTE or LOCAL position, the diesel generator breaker will be automatically tripped by a momentary tripping pulse. The diesel generator will cont,inue running and automatically revert to the isochronous mode. All non-critical protective devices are bypassed.* If a non-critical trip occurs 3I during testing, the diesel generator will trip. On a subsequent
>4 l SIAS or AFAS or LOP the diesel generator will automatically start and run n the isochronous mode The LOCAL control position is selected from the local control panel for diesel generator maintenance testing. A diesel
'I generator LOCAL POSITION alarm will be annunciated in the control room. To prevent any starting of the diesel generator during maintenance, the OFF position is selected at the local c'ontrol panel and a DIESEL GENERATOR INOPERABLE alarm is initiated at the safety equipment status system annunciatory If the preferred power source is lost while paralleled to the diesel generator during testing, the diesel generator will trip on overcurrent and the diesel generator breaker will, trip automatically on a diesel generator shutdown signal. Upon detection of undervoltage on the Class IE 4.16 kV bus, load shedding and sequencing will be ini,tiated as described in section 8.3.1.1.4.6.
L Amendment 14 8.3-44 February 1985
PVNGS FSAR ONSITE POWER SYSTEMS 8.3.1;2.2.17-':Re ulatoi Guide 1.'81--Shared Emer enc and Shutdown Electric S stems for Multi-Unit Nuclear Power Plants.
The requirements of Regulatory Guide 1.81 are met. Each unit has separate and independent onsite ac and dc electric systems capable of supplying minimum ESF loads and loads required for attaining a safe and orderly cold shutdown of the unit assuming a single failure and loss of offsite power. No emergency and shutdown electric systems are shared between units.
J 8.3.1.2.2..18 Re lator Guide 1.89-- uglification of Class IE Equi ment for Nuclear'ower Plants. Regulatory Guide 1.89 endorses IEEE Standard 323-1974,,IEEE Standard for Qualifying
.Class IE Equipment for Nuclear Power Generating Stations.
Comparison of the design with the recommendations of Regulatory Guide 1.89 is discussed in section 1.8.
\
8.3.1.2.2.19 Re lator Guide 1.93-'-Availabilit of Electric Power Sources.
pd(
gpac+ic~Ren f
s) .
'~5/k,SPA The position of Regulatory Guide 1.93 is I<<'
8.3.1.2.2.20 IEEE 308-1974--IEEE Standard Criteria for Class IE Electric S stems for Nuclear Power Generatin Stations.
The Class IE ac power systems are designed to assure that any design basis event, as listed in table 1 of IEEE 308, does not cause the following:
A. A loss of electric power to more than one load group, surveillance devices, or protection system devices sufficient to jeopardize the safety of the unit.
B. A loss of electric power to equipment that could result in a reactor power transient capable of causing significant damage to the fuel or to the reactor cool-ant system.
8.3-72
'ONS'ITE'POWER 'SYSTEMS The Class IE system is capable of "performing its function when subjected Co 'the effects of -any of-the design. basis events.
.The Class .IE loads are designed to perform their functions
. adequately.. for the design variations of voltage and frequency in the Class IE systems.
Controls and indicators for the Class IE 4.16-kV bus supply breakers are .provided in the -control .room and on the switch-gear. Controls and indicators for 'the standby diesel generator power supplies are provided in'he control, room and in the
- ""di;esel"generator control rooms;,
Class IE .equipment and associated design, operating, and main-tenance documents are distinctly identified as described in section 8.3.1.3.
Each type of Class IE equipment is qualified either by analysis, by successful use under similar conditions, by actual test or by a combination of analysis and test to demonstrate its abil-ity to perform its function under applicable design basis events.
Supplementary design criteria of IEEE 308 are addressed in the applicable sections describing specific Class IE equipment.
The surveillance requirements of IEEE 308 are followed in the design, installation, and operation of Class IE systems. *Pre-operational tests are performed in accordance with the pro-are performed as discussed in section 9pQ.Ci A C~+ldYl5 ~
~~
cedures described in section 14.2. Periodic equipment tests
~/+ c4 ~e. ~<+"'~
With regard to Sec<ion 7 of IEEE 308, refer to sections~~<
for operating alternatives under degraded Class IE ac system conditions.
~/q.Q ~g +he Tech"'~~~
Q+ci Rc~+oA5 8.3-73
PVNGS..
' ~
FSAR ONS ITE POWER.-SYSTEMS .-
8.3.2.1.2.2 Batter Char er Ca acit . The capacity of each-Class IE battery charger is based on the largest combined demand of all the steady state loads and the charging current required 'to restore the battery from the design minimum charge state to the fully'charged state within 12 h'ours regardless of the status of the plant during which these demands occur. This is in accordance with Regulatory Guide 1.32.
~ ~ 1 ~ /
8.3.2.1.2.3 .:.Ins ection, Maintenance, and Testin . Testing of.,the dc power- system is, performed prior. to plant operation in.,accordance with IEEE 450-1972 as described in section 14.2.
Subsequent section ~~ tests and inspections will be as described in
'5/+.P.2. o0 +e. Te,+naacp.t Spec>Ac~h<<S.
8.3.2.1.3 Separation and Vent'lation a
The Class IE batteries, chargers, and dc switchgear are located in separate rooms of the Seismic Category I control building, as described in section 8.3.1.1.8.
Each battery room is provided with separate and independent exhaust fans. The ventilation system is designed to preclude the possibility of hydrogen accumulation. Refer to sec-tions 8.3.1.1.8 and 9.4 for details regarding the battery room ventilation system.
8.3.2.2 A~nal sls 8.3.2.F 1 Compliance with Design Criteria and Guides The analysis in this section demonstrates compliance of the Class IE dc power system with General Design Criteria 17 and 18, Regulatory Guides 1.6, 1.22, 1.29, 1.30, 1.32, 1-40, 1.41, 1.47, 1.53, 1.75, 1.81, 1.89, 1.93 and IEEE Standards 308, 323, 344, 383, 384, and 450.
Amendment 3 8.3-96 December 1980
-,PVNGS -FSAR i+aS ITE .-.ROWER -..SYSTENS 8.3.2.2:1.17 .Re lator Guide 1.89-- uglification of Class- IE E i ment for Nuclear Power Plants. Regulatory Guide 1.89 endorses IEEE Standard,323-1974, IEEE Standard for Qualifying Class IE Ec{uipment for Nuclear Power Generating Stations.
.Comparison of the design with the .recommendations of Regulatory Guide 1.89 is discussed in section 1.8.
8.3.2.2.1.18 Re lator Guide 1.93--Availabilit of Electric PowerSources.
p d ( f
'he
'.,'5/p.g position of Regulatory Guide 1.93 is -.
c C +h Td 'i I Sf 'Pi 8.3.2.2.1.19 IEEE Standard 308, 1974--IEEE Standard Criteria for Class IE Electric S stems for Nucleax Power Generatin Stations. The Class IE dc system provides dc electric power to the Class IE dc loads and for control and switching of the Class IE systems. Physical separation, electrical isolation, and redundancy are provided to prevent the occurrence of common failure modes. Design of the Class IE dc system
'includes the following:
A.'he dc system is separated into four subsystems.
B. The -safety actions by each group of loads are independent of the safety actions provided by its redundant counterpart.
C. Each dc subsystem includes power supplies that con-sist of one battery and one battery charger.
D. The batteries are not interconnected.
E. The redundant batteries cannot be made inoperative by a single design basis. event.
Each Class IE distribution circuit is capable of transmitting sufficient energy to start and operate all reguired loads in that circuit. Distribution circuits to redundant equipment are independent of each other. The distribution system is 8.3-101 L
PVNGS FSAR ONSITE POWER SYSTEMS system. Instrumentation is provided. to monitor the status of the battery'charger as follows:
ll I.
"A ='Output voltage at 'the charger and in the control room I
~ ~
B. Output current at the charger and in the control room C. AC and dc breaker position indications at the charger D. Charger malfunction alarm in control room, including input ac undervoltage, dc undervoltage, dc overvoltage,
-'--'--"---'-and output. breaker 'open.
battery charger has an input ac and output dc circuit
'ach breaker. for .isolation of the charger.. Each battery charger power supply is designed to prevent. the'c supply from becom-ing a load on'the battery 'due to a power feedback as the result of the loss of ac power to the chargers. Battery chargers are provided with built-in overvoltage shutdown protection that is capable of tripping the AC input breaker in the event of DC overvoltage.
Equipment of the Class IE dc system is protected and isolated by fuses or circuit breakers in case of short circuit or over-load conditions. Indication is provided to identify equipment that is made unavailable (refer to table 8.3-7).
The Class IE 125V dc subsystem is designed to meet Seismic Category I requirements as stated in section 3.10. The bat-teries, battery chargers, inverters, and other components of dc subsystem are housed in the control building, which is a Seismic Category I structure.
The periodic testing and surveillance requirements for the Class IE batteries are detailed in section M~~ 5/'4 S>>~ ++
Tebnica.l Spec.' canons.
8.3.2.2.1.20 DELETED 8.3.2.2.1.21 IEEE 323-1974--Standard for ualit Class IE Equi ment for Nuclear Power Generatin Stations. Refer to section 8.3.1.2.2.22.
August 1984 8.3-103 Amendment 13
PVNGS FSAR ONSITE POWER SYSTEMS
,8..3.2.2.,1.25. ,IEEE 384-1974 Criteria for Se aration of Class
,IE Equipment and Circuits. *- Refer to section 8.3.1.4 for compliance. p/g.g, Z Spec g caEions).
- 8. 3. 2. 2. 1. 26 DELETED
A. Maintenance and inspections are carried out on a regular scheduled basis to comply with the require-A B.
discussed in section td e.ct &~%As
'~
Performance discharge tests are carried out as d/+ 8 Z. o~ +~e C. T e rating of the battery is approximately 25% greater than that requi ed to supply the emergency load requirements. This margin allows a battery replace-ment criteria of 80~ rated capacity (see sec-tion 8. 3. 2. 1.2. l) .
D. An acceptance test of battery capacity is performed at the factory to determine that it meets the specified discharge rate and duration.
E. Records of the data obtained from inspections and tests are kept along with test procedures to comply with the requirements.
Nhenever any cell's electrolyte level reaches the low level mark, water will be added to increase the level to approximately the midpoint between the high and low electrolyte level marks.
February 1984 8.3-105 Amendment 12
n C 'I PVNGS FSAR WATER SYSTEMS
~ ' ~
funct'ion." ~ ~
C'omponents-'are below gradet with the excep-tion'f e the spray nozzles over the essential spray
~*
ponds. Redundant nozzle systems are available. Each spray pond and its associated train is separated from the other spray pond and its associated train.
9.2.1.8 Tests and Ins ections Preoperational testing is performed in accordance with the test descriptions of section 14.2. Periodic surveillance test-ing. is. described in.section+&-.+~ S/+,7.$ oP the. Tec~~iM QPP cigicckoAs ~
9.2.1.9 Instrument A plication The ESPS instrumentation facilitates automatic operation, remote control, and continuous indication of system parameters (ESP water temperature, ESPS pump flow, ESP inlet flow, pH, ESP water level) both locally and in the control room.
Controls and instrumentation necessary for operation of the ESPS pumps are located in the control room. Local instrumen-tation and controls also are provided at ESPS pumps and ECWS heat exchangers for mainterance, testing, and performance evaluation.
Specifically, control room process indication and alarm is provided to enable the operator to evaluate the ESPS perform-ance and to detect malfunctions. ESPS pump discharge pressure is displayed and alarmed to detect an abnormally low pressure (pump failure, piping break) or abnormally high pressure (piping blockage, closed valves). Spray pond levels are indicated to show a low or high level and'emperatures condition or a high temperature condition in a spray pond.
Control conditions of level and temperatures are also alarmed in the control room. The ESPS water discharge temperatures from the ECWS heat exchangers are indicated in the control room. A high temperature condition is 9.2-14
the fuel pool heat exchangers must be supplied cooling water by'the 'essential cooling water system (ECWS)'.'ach fuel pool-heat exchanger is supplied water .
separately by each train of the ECWS; i.e., one heat exchanger by train A and the other heat exchanger by train B:.
Valves associated with 'switching service from the NCWS to the ECWS are manually operated, Seismic I, and Safety Class 3. These valves are also used to isolate the ECWS from the NCWS. They are located in the auxiliary 'building.
Sufficient l
time would be available for the operator to accessthe auxiliary building to manually actuate these valves since the fuel pool does not require continuous cooling.
Safety Evaluation Ten Components of the ECWS are located such that missiles h
from any source would not impair the system's func-tional requirements. The two trains of the ECWS are physically separated and are routed such as to be protected from missiles that could be potentially generated from other sources. Refer to section 3.5 for a discussion of missile protection.
9.2.2.1.8 Tests and Inspections Preoperational testing is performed in accordance with the testing is described in section Te~hn' Sp c P ~4o~a.
~~
test descriptions of section 14.2. Periodic surveillance H/'t.7'3 o0 &e.
9.2.2.1.9 Instrument Application Refer to section 9.2.1.9 for a presentation of the ECWS interfaces to the ESPS.
9.2-29
PIGS.-'SAR MATER" SYSTEMS=
9.2.5.5 :-Tes'ts and I'ns ections:--"."."
Refer to 'se'ction 1'4 2 for 'a"discussi'on" of'p'reoperat'ional test procedures. Refer to section for a description of periodic surveillance testing.
~/R.V.$ ~g A~ Ted,n.& Qe~i~~~~'.2.5.6
--Instrumentation A lications The water level in the ESP is monitored continuously so that there is always sufficient water to 'ensure the continuous capability of the ESP to perform its safety functions. The water temperatures of the ESP are also monitored.
8 ) Amendment 8 9.2-94D March 1982
'PVNGS-'FSAR
'WA'TER'S'tSTEMS useable water is reserved. This ensures a sufficient supply to the auxiliary feedwater pumps. A separate line is connected to the tank at a lower elevation to supply the auxiliary feedwater pumps with the reserved water supply. A single active failure analysis for the condensate storage facility is provided in table 9.2-23.
-:9;2.6.4 'ESSAR" Interface'Evaluation Refer to sections 5.1.5 and 9.3.4.2.
9.2.6.5 Tests and Ins ections Preoperational testing is performed in accordance with the test.
descriptions of section 14.2. Periodic surveillance testing is described in section ~~~ 5/1 7 o0 +be. 'Technical <Fec'~'ca'Iiens-1 9.2.6.6 Instrumentation A vlications A flow transmitter with output to the computer is provided on the condensate tank fill line.
A level detection system is installed on the condensate tank with level signals transmitted to the automatic tank level controller. Level indication is provided locally and in the control room. Low and high level alarms are provided in the control room. The low level alarms annunciate when the tank volume falls below 530,000 gallons, 330,000 gallons and 20,000 gallons.
9.2.7 SHUTDOWN COOLING SYSTEM Refer to section 5.4.7.
9.2.8 TURBINE COOLING WATER SYSTEM Tne turbine cooling water system (TCWS) provides cooling for the nonnuclear related components in the various turbine plant 9.2-104
PVNGS. FSAR WATER SYSTEMS 9.2.9.2.4 Tests and Inspections Preoperational testing, is, performed in accordance with the test
~~
I decriptions of section 14.2. Periodic surveillance testing is described in secticn 3/g 7.( c0 +he Technical QeciR'ceh<<S.
s
- 9. 2. 9. 2. 5 Instrumentation Applications The chiller units and chilled water pumps for the essential chilled water system are automatically actuated upon receiving
'ny of the signals shown on figure 9.2-10.
After automatic startup of the two essential chilled water trains, train A and;train B, the operator has manual override capability on the essential trains. The operator is able to determine which train he wishes to remain in operation, which train he wishes to deactivate, and can reactivate the standby train manually, as required.
A temperature and capacity controller is provided with each essential chiller unit and maintains a constant. chilled water supply temperature when the unit is working. Integral flow switches prevent the chiller from operating unless there is cooling water flow through the condenser and chilled water flow in the evaporator. A trip of any chiller or pump causes an alarm and the operator puts the appropriate standby unit into service. Essential chilled water system'ifferential pressure indication and alarm and essential chiller outlet temperature indication and alarm are provided in the control room to monitor system operation and efficiency. Additional local display instrumentation and test points are placed in the equipment, areas for periodic checkout of the system.
The essential chilled water expansion tank is provided with level gage glass to show low or high level condition in the closed loop. Critical conditions of the tank level and pres-sure are alarmed in the control room for leak detection.
9 '-124
~
PVNGS FSAR PROCESS AUXILIARIES P1$
1 G. Thermal Limitations The ventilation systems are designed in accordance with CESSAR Section 3.11 to maintain the ambient conditions in the auxiliary building between 50 and 104F, and in the containment building between 50 and 120F, under normal operating conditions (refer to section 9.4).
- 2. Following a loss-of-coolant accident, including the subsequent recirculation mode of operation,
'. the ambient air conditions of the CVCS equipment located in the auxiliary building are controlled in accordance with the requirements of section 3.11.
H. Monitoring Not applicable.
I. Operational controls Not applicable.
Inspection and Testing
K. Chemistry/Sampling T8clM cJ p/y [ 'Z gg ~~
Not applicable. g p~~t QcP4 o AS L. Materials
- 1. The insulation used on austenitic stainless steel is discussed in section 5.2.3. Cleaning and contamination procedures are also discussed in section 5.2.3. Conformance to Regulatory Guides 1.36 and 1.37 is discussed in sections 6.1 and 1.8, respectively.
9 '-86
' 'i PVNGS FSAR OTHER AUXILIARY SYSTEMS I ~
~,capable,.of measuring 150- of rated capacity of the largest fire pump.
4 ~
Fire detection instrumentation will be tested and inspected at regular intervals as necessary to maintain highly reliable systems. The regular intervals of tion ~~
testing and inspection will be as described in sec-
>/Q.'3 3 o0 e
6e Technice-( +pec'Fice'hc">
9.5.1.5 Personnel Qualification and Trainin 9.5.1.5.1 Overall Requirements of the P'PiIGS Fire Protection Program Ultimate responsibility for the overall fire protection program at PVNGS rests with the Director of Nuclear Operations. The responsibility for formulation and assurance of program implemen-tation has been delegated to staff personnel having training and experience in fire protection and nuclear plant safety. The fire protection program provides:
A. Xnspection and maintenance of fire protection equipment.
B. An organization to fight fires and deal with related emergencies when they occur.
C. Determination that fire apparatus. functions properly and that procedures and practices are in accordance with accepted rules and regulations.
D. Maintenance of records on fire protection equipment.
E. Fire fighting training for station employees.
Xnspection of the station and control of transient fire loads.
G. Fire investigation and reporting including review of fire incidence arid corrective action.
August 1984 9.5-35 Amendment 13
PVNGS FSAR OTHER AUXILIARY SYSTEMS 9.5.4.5 Ins ection and Testin Requirements The diesel'enerator fuel oil storage tank for each diesel generator is tested by nondestructive methods in accordance with ASME Boiler and Pressure Vessel Code, Section III, Class 3 and is subjected to routine tests and inspections during construction and installation.
Secbo~ '3/g.a.l o4 Refer to>the technical specifications for operational tests and inspections.
9.5.4.6 . Instrumentation Ao lications A pressure switch, installed on the transfer pump discharge initiates an alarm in the control room and local diesel gen-erator control panel if the day tank level is low and low pressure. exists in this header. The alarm indcates that fuel.oil is not being pumped to the day.tank.
Level switches on each day tank start or stop the transfer pump at preset level points. Level switches also initiate low-low day tank level alarms in the control room and diesel generator panel in the generator room. Refer to section 7.4 for the DGFOS fuel oi'ransfer logic.
9.5.5 DIESEL GENERATOR COOLING WATER SYSTEM The diesel generator cooling water system '(DGCWS) removes the waste heat of combustion from the diesel engine. Each engine is provided with an independent. DGCWS, and the description which follows applies to each system.
May 1981 9.5-63 Amendment 4
. PVNGS FSAR g.pz af He,7echnl'ca/ gee'h'c~6< and CESSAR- Section 16.3.4.8 will be made to the NRC. A program of fuel assembly sipping will be conducted ai the next refue'ling to determine the location and number of failed fuel rods, if a significant number is indicated.
QUESTION 9A.41 (NRC Question 410.7) (9.3.1)
Concerning the compressed air system, provide the following additional information:
Describe the means provided to verify that proper instrument air quality will be maintained over the plant life to assure the safety function of the system (i.e., air operated valves will fail in their safe position on loss of instrument air supply). Include the air quality limits which should not be exceeded in order to assure the above safety function.
b) Verify that a single failure of any air operated valve to assume its fail safe position will not prevent the function of a safety-related system or compromise the ability to safely shut down.
RESPONSE
a) In the PVNGS design the instrument air system is .not a safety-related system. However, local annunciation is provided for the following changes in instrument air quality:
(1) High differentia'ressure across'he prefilter (2) High differential pressure across the dryer October 1981 9A-31 Amendment 6
PVNGS FSAR MAIN STEAM SUPPLY SYSTEM H. Safety Evaluation Eight.
A branch connection upstream of the MSIVs from each steam generator provides steam to operate the auxiliary feedwater pump turbine. Refer to section 10.4.9.
10.3.3.2 CESSAR Interface Evaluation Refer to section 5.1.5.
l0.3.4 INSPECTION AND TESTING REQUIREMENTS Refer to section 14.2 for pre-operational testing requirements.
Refer to section 3.9> and>
~
~ for inservice testing and inspection requirements.
~
~
g~~yo~ p/q,g I o0 +e.'Tech~'icnl 5)~c 0
~ ~
10.3;5 - WATER CHEMISTRY (PWR) 10.3.5.1 Chemistr Control Basis Refer to CESSAR Section 10.3.4.1.
10.3.5.2 . Corrosion Control Effectiveness Refer to CESSAR Section 10.3.4.2.
10.3.5.3 Chemistr Control Effects on Iodine Partitionin Refer to CESSAR Section 10.3.4.3. The partition factor assumed for the condenser vacuum pump outlet is discussed in sec-tion 11.1.8.
10.3.6 STEAM AND FEEDWATER SYSTEM MATERIALS l0.3.6.1 Fracture Tou hness The materials are in compliance with the ASME Boiler and Pres-sure Vessel Code, Sections II and III, 1974 Edition through the Winter, l975 Addenda. The fracture toughness properties meet 10.3-16
REVIEW AND AUDIT 13.4.3 AUDIT PROGRAM t
A comprehensive program of planned and documented audits is carried out to verify compliance with, and effectiveness of, implementation of the administrative controls and Quality Assurance (QA) program and to assist the NSG in the execution of its responsibility for independent review of operating activities that affect nuclear safety.
Audits are performed in accordance with approved procedures.
The frequency and scope of the audits is discussed in sec-(.E,5 < W wl *<a> P...,
Audit assignments are such that the Audit Team members will not perform audits of activities for which they have immediate responsibility.
Written reports of audits are reviewed by the NSG and by appro-priate members of management, including those having responsi-bility in the area audited. Appropriate and timely followup action, including re-audit of deficient areas as appropriate, is taken to ensure overall effectiveness of the review and audit program.
The QA audit. program for operations is discussed in section 17.2.
August 1984 13.4-3/-4 Deleted Amendment 13
~,
PLANT PROCEDURES 13.5.1.3 Procedures The following are descriptions of administrative procedures that will be prepared for PVNGS:
,... A. Procedures for Shift Supervisors and Operators
- 1. Senior reactor operator's authority and responsibilities
- a. Describes senior reactor operator's duties, responsibilities, and authority.
- 2. Reactor operator's authority and responsibilities
- a. Describes the reactor operator's duties, responsibilities, and authority.
/.2. o4 &~ &echoic
- 3. Conduct of ope ations Q pec<Qccl:4~&
a ~ Procedures are wri en to implement the pro-visions of sectionz~~concerning licensed personnel on shift. These procedures will include provisions of 10CFR50.54 (i) through (m).
- b. The "at the controls" area o the control room is shown in figure 7.5-1.
B. Special Orders of a Transient or Self Cancell'ing Nature Special orders will be written to issue instructions which have short term applicability and which require dissemination.. These orders will be reviewed on at least an annual basis for the purpose of purging and updating.
C. Equipment Control Procedures Equipment control procedures are written to provide control over the status of station equipment, of pur-chased material, and of nonconforming material. Such procedures will include:
- 1. Work authorization.
Amendment 8 13.5-2 March 1982
PVNGS FSAR APPENDIX lsB 14B.14 REACTOR CONTAINMENT INTEGRATED AND LOCAL '~LEAK 'RATE';
TESTS
1.0 'BJECTIVE 4
To demonstrate, prior to initial reactor
'operation, that. leakage through the primary reactor containment and systems and components penetrating primary containment do not exceed the allowable leakage rate values as specified in the Technical Specifications.
2.1 Construction activities completed.
2.2 Structural integrity test (described in subsec-tion 3.8.1.7) satisfactorily completed.
2.3 Leakage rate determination instrumentation avail-able and properly calibrated.
- 2. 4 Containment ventilation system, personnel airlock and isolation valves are operable.
2.5 Containment inspection completed.
3.0 TEST METHOD 3.1 Perform individual local leakage tests on con-tainment isolation valves and penetrations as described in Section 6.2.
3.2 Perform a containment building integrated leakage rate test at the calculated peak internal pres-sure per Section 6.2.
4.0 ACCEPTANCE CRITERIA 4.1 The containment leakage shall not exceed the plant technical specification limits stipulated in Sed ~ S/4.(..i o4 T~akn<ccaf Space 9 akim~ ~
Amendment 9 14B-24/-24A and -24B Deleted August 1982
Pi