Text

Forwards marked-up Final Draft Tech Specs,Incorporating Changes & Justifications for Changes,Per 851018 Request
	ML17299A717
Person / Time
Site:	Palo Verde
Issue date:	11/01/1985
From:	Van Brunt E; ARIZONA PUBLIC SERVICE CO. (FORMERLY ARIZONA NUCLEAR
To:	Knighton G; Office of Nuclear Reactor Regulation
References
	ANPP-33898-EEVB, TAC-60706, NUDOCS 8511070179
	Download: ML17299A717 (197)
	v • d • e

REGULATORY INFORMATION DISTRIBUTION TEM (RIDS)

II I

ACCESSION NBR:8511070179 DOC ~ DATE: 85/11/01 NOTARIZED:

NO DOCKET 0'ACIL'.STN 50-529 Palo Verde Nuclear Stationi Unit 2R Arizona Publ i 05000529 AUTH BYNAME AUTHOR AFFILIATION VAN BRUNTRE;ED Arizona Nuclear Power Project (formerly Arizona Public"Serv RECIP,NAME RECIPIENT AFFILIATION KNIGHTONiG AN ~

Office. of Nuclear Reactor Regulationp Director

SUBJECT:

Forwards marked-up final draft Tech SpecsRincorporating changes 8 justifications for changesiper 851018 request

~ A+i DISTRIBUTION CODE:

B0010 COPIES RECEIVED:LTR ENCL SIZE:

TITLE: Licensing Submittal:

PSAR/FSAR Amdts 8 Related Corr espondence NOTES:Standardized plant.

05000529

- RECIPIENT IO CODE/NAME NRR/OL/AOL NRR LB3 LA INTERNAL; ACRS-91 ELD/HOS3 IE/DEPER/EPB 36 NRR ROERM ~ L NRR/DE/CEB 11 NRR/OE/EQB 13 NRR/DE/MEB 18 NRR/DE/SAB 24 NRR/OHFS/HFEBPO NRR/DHFS/PSRB NRR/DS I/AEB 26 NRR/DS I/CPB 10 NRR/DSI/ICSB 16 NRR/DSI/PSB 19 NRR/DSI/RSB 23 RGN5 EXTERNAL; 24X DMB/DSS (AMDTS)

NRC. POR 02 PNL GRUEL"RR COPIES LTTR ENCL 6

1-1 1

1 1

1=-

1 1

1 31 1

1~

1Cr 0

RECIPIENT IO CODE/NAME NRR LB3 BC LICITRA E, 01 AOM/LFHB IE FILE:

IE/DQAVT'/QAB21 NRR/OE/AEAB NRR/DE/EHEB NRR/DF/GB 28 NRR/OE/HTEB 1.7 NRR/DE/SGEB 25 NRR/DHFS/LQB 32 NRR/DL/SSPB NRR/DS I/ASB NRR/OSI/CSB 09 NRR/DS I/h1ETB 12 NRR/DSI RAB 22' FIL 04

/MI8 BNL(AHDTS ONLY)

LPOR 03 NSIC-05 COPIES LTTR ENCL" 1

0 1

14' 1

1 1

2 1

1 1

1:

1 1

1 1$

1 0

1 1

15 TOTAL NUMBER OF COPIES REQUIRED:

LTTR 52 ENCL" ~N

l A,

E r

41 C

V l

I' Jl 41 ~

~ '0

Arizona Nuciear Power Project P.O. BOX 52034

~

PHOENIX, ARIZONA55072-2034 ANPP-33898-EEVB/JRP November 1, 1985 Director of Nuclear Reactor Regulation Attention:

Mr. George W. Knighton, Project Director PWR Project Directorate ¹7 Division of Pressurized Water Reactor Licensing B U.S. Nuclear Regulatory Commission Washington, D. C.

20555 r.

Subject:

Palo Verde Nuclear Generating Station Unit 2 Docket No. 50-529 Plant Technical Specifications File: 85-056-026

Reference:

Letter from G.

W. Knighton, NRR, to E. E.

Van Brunt, Jr.,

ANPP, dated October 18, 1985

Dear Mr. Knighton:

The referenced letter transmitted to ANPP the "Final Draft" copy of PVNGS Unit 2 Technical Specifications.

After our review of the final draft, we have determined that the enclosed changes need to be incorporated in this document.

We understand that this corrected document will be the basis for our certifying the Technical Specifications for Unit 2.

As requested in the referenced letter, attached please find changes and justification for these changes for incorporation into the Unit 2 Technical Specifications.

Should you have any questions or require additional information, please call Mr.

W. F. Quinn at (602) 943-7200, Extension 4087.

Very truly yours E. E.

Van Brunt, Jr.

Executive Vice President Project Director EEVB/JRP/llg Attachment cc: Director, Region V, USNRC NRC Project Manager, E. A. Licitra NRC Resident Inspector, R. P.

Zimmerman NRC Tech.

Spec.

Reviewer, S.

Brown 8511070179 851101

'DR ADOCK 05000529 PDR

Proposed Change and Reason for Change to the Basis for Technical Specification 3/4.3.3.3 The proposed change is as follows.

Add the following sentence to the end of the existing.basis statement on page B 3/4 3-3:

"The seismic instrumentation, for the site is located in Unit 1 at the locations indicated in table 3.3-7."

The reason for this change is as follows.

Table 3.3-7 in the Unit 2 Technical Specifications is identical to the table in the Unit 1 Technical Specifications.

The addition is to clarify the fact that there is only one set of seismic instruments for the site and that they are in Unit 1.

1'

t

, 8511070179, INSTRUMENTATION BASES

Response

time may be demonstrated by any series of sequential, overlapping, or total channel test measurements provided that such tests demonstrate the total channel response time as defined.

Sensor response time verUication may be demonstrated by either (1) in place, onsite, or offsite test measurements or (2) utilizing replacement sensors with certified response times.

3/4. 3. 3 MONITORING INSTRUMENTATION 3/4. 3. 3. 1 RADIATION MONITORING INSTRUMENTATION The OPERABILITY of the radiation monitoring channels ensures that:

(1) the radiation levels are continually measured in the areas served by the individual channels and (2) the alarm or'automatic action is initiated when the radiation level trjp setpoint is exceeded.

3/4.3.3.2 IHCORE DETECTORS The OPERABIlITY of the incore detectors with the specified minimum comple-ment of equipment ensures that the measurements obtained from use of this system accurately represent the spatial neutron flux distribution of the reactor core.

3/4. 3. 3. 3 SEISMIC INSTRUMENTATION The OPERABILITY of the seismic instrumentation ensures that sufficient capability is available to promptly determine the magnitude of a seismic event and evaluate the response of those features important to safety.

This capabil-

. ity is required to permit comparison of the measured response. to that used in the design basis for the facility to determine if plant shutdown is required pursuant to Appendix A of 10 CFR Part'00.

The instrumentation is consistent with the recommendations of Regulatory Guide 1. 12, "Instrumentation for Earth-quakes," April 1974 as identified in the PYNGS FSAR. -rate. s~ia~ic i~~W~~~ka%ia~

pa~ 4m ~4~ ia %ac.aug g~ ~z ~~ ~p-v,

'/4.3.3. 4 METEOROLOGICAL INSTRUMENTATION The OPERABILITY of the meteorological instrumentation ensures that suffi-cient meteorological data are available for estimating potential radiation doses to the public as a result of routine or accidental.release of radioactive materials to the atmosphere.

This capability is required to evaluate the need for initiating protective measures to protect the health and safety of the public and is consistent with the recommendations of Regulatory Guide 1.23 "Onsite Meteorological Programs,"

February 1972.

Mind speeds less than 0."6 HPH cannot be measured by the meteorological instrumentation.

3/4.3.3.5 REMOTE SHUTDOMN SYSTEM

'The OPERABILITY of the remote shutdown system ensures that sufficient capability is available to permit safe shutdown and maintenance of HOT STANDBY of the facility from locations outside of the control room.

This capability is required in the event control room habitability is lost and is consistent with General Design Criterion 19 of 10 CFR Part 50.

PALO YERDE - UNIT 2 B 3/4 3"3 OCT1l 1985

C

Proposed Changes and Basis for Changes to Technical Specification 3/4 11.2.5 The proposed changes are on the attached marked up page.

The basis for the change is as follows.

The change is requested to make this surveillance requirement consistent with the requirements of specification 3.3.3.9 table 3.3-13, which allows for grab sampling if the continuous monitoring instrumentation is out of service.

l 0

RADIOACTIVE EFFLUENTS EXPLOSIVE GAS MIXTURE LIMITING CONDITION FOR OPERATION

.3.11.2.5 The concentration of oxygen in the waste gas holdup system shall be limited to less than or equal to 2X by volume whenever the hydrogen concentration exceeds 4X by volume.

APPLICABILITY: At al 1 times.

ACTION:

Mith the concentration of oxygen in the waste gas holdup system greater than 2X by volume but less than or equal to 4X by volume, reduce the oxygen concentration to the above limit within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months .

b.

C.

With the concentration of oxygen in the waste gas holdup system greater than 4X by volume, immediately suspend all additions of waste gases to the system and reduce the concentration of oxygen to 'less than 4X by volume within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE RE UIREMENTS

4. 11.2.5 The concentration of hydrogen or oxygen in the waste gas holdup 1

1 1

1 1i1 y~

monitoring the waste gases in the waste gas holdup system

. Specification 3.3.3.9.

(p/ Qggo+QAgcE ca1 PH PALO VERDE - UNIT 2 3/4 11"14

C f

Proposed Changes and Basis for Changes to Technical Specification 3/4 8.1.1 The proposed changes are on the attached marked up pages.

The basis for the changes are as follows.

Generator output breaker control logic does not rely on the diesel generator reaching 600 RPM, but uses voltage and frequency reaching a

specific value before the breaker will close.

Since achieving the required voltage and frequency indicates that proper engine speed has been attained, we are proposing that references to speed in 4.8.1.1.2.a.4 and 4.8.1.1.2.e be deleted.

The present Unit 1 Technical Specification for performance of the diesel generator test in 4.8.1.1.2.a.4 allows the option of starting diesels by one of four methods.

The Unit 2 Specification 4.8.1.1.2.c stipulates that diesel must be started on a staggered test basis by one of three methods.

The three test methods required cause the HSF busses and emergency system to be placed. in a degraded mode while the test is being performed.

The tests place certain emergency systems in operation *and cause the diesel to be placed in the emergency mode.

We do not believe the intent of surveillance testing is to place systems in a degraded mode, therefore we have proposed the addition of a manual start and deletion of the staggered test basis.

We also note that other utilities, i.e. Waterford 3, San Onofre 2 8 3, and Byron station do not use a staggered test choice to start their diesels.

'he diesel generator and emergency generation system is fully tested every 18 months proving that the diesels will start and load under emergency conditions.

Presently the Unit 1 Technical Specifications do not require that the diesels be started and loaded under these conditions more than every 18 months.

We do not believe these technical specifications should impose any more severe restrictions than Unit 1.

Insert A

The diesel generator shall be started for this test by using one of the following signals:

a)

Manual b) 'imulated loss-of-offsite power by itself c)

Simulated loss-of-offsite power in conjunction with an ESF actuation test signal.

d)

An ESF actuation test, signal by itself Insert B

With generator voltage and frequency at 4160+420 volts and 60+1.2 Hz

El ECTRICAL POMER SYSTEMS SURVEILLANCE RE UIREMENTS 4.8. 1. 1.1 Each of the above required physically independent circuits between the offsite transmission network and the onsite Class 1E distribution system shall be:

a.

Determined OPERABLE at least once per 7 days by verifying correct breaker alignment indicating power availability b.

Demonstrated OPERABLE at least once per l8 months during shutdown by manually transferring the onsite Class 1E power supply from the normal circuit to the alternate circuit.

4.8.1.1.2 Each diesel generator shall be demonstrated OPERABLE:

a.

In accordance with the frequency specified in Table 4.8-1 on a STAGGERED TEST BASIS by:

1.

Verifying the fuel level in the day tank.

2.

Verifying the fuel level in the fuel storage tank.

3.

4.

5.

Verifying the fuel transfer pump can be started and transfers fuel from the Storage system to the day tank.

Verifying the diesel generator can start*" and accelerate -to-

. with generator voltage and frequency at 4160 2 420 volts and 60 ~ 1.2 Hz in less than or equal to

.10 seconds.

Subsequently, verifying the generator is synchronized, gradually loaded"" to an indicated 5200-5400 kM""" and operates for at least 60 minutes.

lhlS E4T Verifying the diesel generator is aligned to provide standby power to the associated emergency busses.

b.

C.

At least once per 92 days by verifying that a sample of diesel fuel from the fuel storage tank obtained in accordance with ASTM-D270-65, is within the acceptable limits specified in Table 1 of ASTM D975-74 when checked for viscosity, water and sediment.

At least once per 184 days the diesel generator shall be started""

and accelerated to at least 600 rpm in less than or equal to 10 seconds.

The generator voltage and frequency shall be 4160 i 420 volts and 60 t l.2 Hz within 10 seconds after the start signal.

I

""This test shall be conducted in accordance with the manufacturer's recommen-dations regarding engine prelube and warmup procedures, and as applicable regarding loading recommendations.

"""This band is meant as guidance to avoid routine overloading of the engine.

Loads in excess of this band for special testing under direct monitoring of the manufacturer or momentary variations due to changing bus loads shall not invalidate the test.

PALO VERDE - UNIT 2 1

3/4 8-3 OCTA) )g

0

ELECTRICAL POWER SYSTEM SURVEILLANCE RE UIREHENTS Continued

.8.1. 1. 2 (Continued) 1 The generator shall be mariually synchronized to its appropriate emergency

bus, loaded to an indicated 5200-5400""*

kW in less than or equal to 60 seconds, and operate for at least 60 minutes.

The diesel generator shall be started for this test by using one of the following signals on a STAGGERED TEST BASIS:

a)

Simulated loss of offsite power b)

Simulated loss of offsite power ESF actuation test signal.

c)

An ESF actuation test signal by by itself.

in conjunction with an itsel f.

This test, if it is performed so it coincides with the testing required by Surveillance Requirement 4.8. 1.1.2.a.4, may also serve to concurrently meet those requirements as well.

d.

At least once per 18 months during shutdown by:

3..

Subjecting the diesel to an inspection in accordance with procedures prepared in conjunction with its manufacturer's recommendations for this class of standby service.

2.

Verifying the generator capability to reject a single largest load of greater than or equal to 839 kM (Train 8 AFW pump) for emergency diesel generator B or 696 kW for emergency diesel generator A (Train A HPSI pump) while maintaining voltage at 4160 t 420 volts and frequency at 60 L 1.2 Hz.

3.

Verifying that the automatic load sequencers are OPERABLE with the interval between each load block within t 1 second of its.

design interval.

4.

Simulating a loss of offsite power by itself, and:

a)

Verifying deenergization of the emergency busses and load shedding from the emergency busses.

b)

Verifying the diesel starts"" on the auto-start signal,

'nergizes the emergency busses with permanently connected loads within 10 seconds, energizes the auto-connected shut-down loads through the load sequencer and operates for greater than or equal to 5 minutes while its generator is

""This test shall be conducted in accordance with the manufacturer's recommen-dations regarding en'gine prelube and warmup procedures, and as applicable regarding loading recommendations.

<<"This band i,s meant as guidance to avoid routine overloading of the engine.

Loads in excess of-this band for special testing under direct monitoring of the manufacturer or momentary variations due to changing bus loads shall not ihvalidate the test.

PALO VERDE " UNIT 2 3/4 8-4

0

ELECTRICAL POMER SYSTEMS SURVEILLANCE REQUIREMENTS Continued Verifying that the auto-connected loads to each diesel generator do not exceed the continuous rating of 5500 kM.

Verifying the diesel generator's capability to:

a) b)

c)

Synchronize with the offsite power source while the generator is loaded with its emergency loads upon a

simulated restoration of offsite power, Transfer its loads to the offsite power source, and Proceed through its shutdown sequence.

10.

Verifying that the following diesel generator lockout features prevent diesel generator starting only when required:

a) turning gear-engaged

- b) emergency stop e.

At least once per 10 years or. after any modifications which could affect diesel generator interdependence by starting*" both. diesel generators simultaneously, during shutdown, and verifying that both diese1 generators accelerate in 1ess than or equal to 10 seconds.

i>>dsiti 4.8.1.1.3

~Re orts - All diesel generator failures, valid or nonvalid, shall be reported to the Commission within 30 days in a Special Report pursuant. to Specification 6.9.2.

Reports of diesel generator failures shall include the information recommended in Regulatory Position C.3.b of Regulatory Guide 1.108, Revision 1, August 19?7.

If the number of failures in the last 100 valid tests (on a per nuclear unit basis) is greater than or equal to 7, the report shall be supplemented to include the additional information recommended in Regulatory Position C.3.b of Regulatory Guide 1.108, Revision 1, August 1977.

gq~This test shall be conducted in accordance with the manufacturer's recommen-dations regarding engine prelube and warmup procedures, and as applicable regarding loading recommendations' PALO VERDE - UNIT 2 3/4 8"6 OCT g g fggg

0

Proposed Change and Reason for Change to Technical Specification Table 3.3-4 The proposed changes are on the attached marked up page.

The basis for the change is as follows:

Table Notations (1) and (3), Nodes 3-4.

This is being changed for consistency; CESSAR is also being changed as marked up.

0

TABLE 3. 3-4 (Conti nued}

Pf Pg minimum of 100 sia (4) I of the distance between steam generator upper and lower level wide range instrument nozzles.

TABLE NOTAT'IONS (3)

In NDDES 3-'li value may be decreased manually, to a

p as pressurizer pressure is reduced, provided the margin between the pres-surizer pressure and this value is maintained at less than or equal to 400 psi; the setpoint shall be increased automatically as pressurizer pressure is increased until the trip setpoint is reached.

Trip may be manually bypassed below 400 psia; bypass shall be automatically removed whenever pressurizer pressure is.greater than or equal to 500 psia.

(2) I of the distance between steam generator upper and lower level narrow range instrument nozzles.

(3)

In NDDES 3-g value may be decreased manually as steam generator pressure is reduced, provided the margin between the steam generator pressure and this value is maintained at less than or equal to 200 psi; the setpoint shall be increased automatically as.steam generator pressure is increased until the trip setpoint is reached.

PALO VERDE - UNIT 2 3/4 3-27 OCT 11 ](8"

~

Proposed Change and Reason for Change to Technical Specification Table B 3/4.4-1 The proposed changes are on the attached marked up page.

a)

The basis for the change is as follows:

Deletion of the double asterisk ("') in the footnote and for the Outlet Nozzle temperature of Charpy V-Notch.

This change is requested based on information from Combustion Engineering (C-E) that the values listed are the lowest values for 2 or,3 tests.

In the Unit 1 Technical Specifications, a vendor supplied (C.-E) with the average test va3.ue.

Therefore there is a difference between Units 1 and 2.

b)

Deletion of the term:

FOR LONGITUDINALDIRECTION The reactor vessel toughness (forgings) in the attached Table. are in accordance with Appendix G to 10 CFR Part 50.

'j i

'0 I I'I I

128-201 128"201 128"201 128-201 131"102 131-102 128"301 128-301 131-101 131"101 131-101 131,-101 126-101 106-101 F-774-01 F"774-02 F-774-03 F"774-04 F-767-01 F"767-02 F"?64-01 F-764-02 F-766-01 F-766-02 F"766-03 F-766-04 F-762-01 F-761-01 SA 508"CL3 SA 508-CL3 SA 508-CL3 SA 508"CL3 SA 508-CLl SA 508-CL1 SA 508-CL2 SA 508"CL2 SA 508-CL1 SA 508-CL1 SA 508"CLl SA 508-CL1 SA 508-CL2 SA 508-CL2 PIECE HO.

CODE NO.

MATERIAL TABLE B 3/4.4-1 REACTOR VESSEL TOUGHNESS

~FDRG INES DROP..

MEIGHT RESULTS VESSEL LOCATION

~F Inlet Nozzle

-20 Inlet Nozzle "30 Inlet Nozzle

-40 Inlet Nozzle

-40 Outlet Nozzle Safe End

-30 Outlet Nozzle Safe End

-30 Outlet Nozzle

-10 Outlet Nozzle

'10 Inlet Hozzle Safe End

-10 Inlet Nozzle Safe End 0

Inlet Nozzle Safe End

-30 Inlet Nozzle Safe End

-30 Vessel Flange "40 Closure Head Flange

-50 RTNDT(b)

~op "20

-30

".40

-10

-10 "10

-1P

+10

+10 "20

-40

-50 N.A..

N.A.

H.A.

u Q

+

+ +%5

+

+30 o

+

Owxo 8 +So 2 437 8 bAg

+30 5IR* 0 5**cs 0 yy

+ 0+z.q 7 is.'q 7% Qg

- G-SL.

" 0-g(

+(8 +

%5'0

$ 0IRITR,4-34

+f5 s

+(0 + 5g 0

"$8 y 2g "4

H.A.

N.A.

H.A.

N. A.

N.A.

TEMPERATURE OF MINIMUM UPPER, CHARPY V-HOTCH" SHELF C

ENERGY 9 30 9 50 MNAL ft - lb ft " lb 4@KGR8H;ft lb

+ 0-'%

0 +i~

N.A.

N.A. = Not Applicable (no minimum upper shelf requirement).

= LOWer bOund CurVe ValueS< ci-ci e~<gczy -g p~yv.ioo.g.

~ u (a)

= Determined per applicable ASME-BPV-Code Sect. III, Subsection HB, Article NB-2331-(a-l,2,3).

(b)

= 0'nd 180'pecimens had the same values.

.;7>

0

TABLE B 3/4.4-1 (Continued)

REACTOR YESSEL TOUGHNESS (PLATES)

PIECE HO.

CODE NO.

MATERIAL YESSEL LOCATION DROP TEMPERATURE OF MINIMUM UPPER WEIGHT RT (a)

CHARPY V-NOTCH*

SHELF C

ENERGY RESULTS NDT 8 30 9 50 FOfMONKRUBIHAL

~F)

~F ft lb f 1b

%IMHSN f 142-102 142-102 142-102 124"102 124-102 124-102 122-102 122-102 122-102 102-102 102-102 150-102 150-102 F-773-01 F-773-02 F-773-03 F-765-04 F-765-05 F-765-06 F-765-01 F-765-02 F-765-03 F-770-01 F-770-02 F-771-01 F"771-02 SA 533-GRB-CL1 SA 533-GRB-CLl SA 533-GRB-CL1 SA 533-GRB-CL1 SA 533-GRB-CL1 SA 533-GRB-CL1 SA 533-GRB-CL1 SA 533-GRB-CL1 SA 533-GRB-CL1 SA 533-GRB-CLl SA 533-GRB-CL1 SA 533-GRB-CLl SA 533-GRB-CL1 Lower Shell Plate Lower Shell Plate Lower Shell Plate Intermed.

Shell Plate Intermed.

Shell Plate

'ntermed.

Shell Plate Upper Shell Plate Upper Shell Plate Upper Shell Plate Closure Head Dome Closure

)lead Dome Bottom Head Dome Bottom Head Dome "40

-50

-60

-30

-20 "30

-30

-40

-30

-60.

-50 "90 "70

+10 0

-60

-20

+10

+10 0

-20

-40 "50

-50 6~z<

+4 4-c~

7-sz.

+14 4 viz.

+90 6 vi5

+$8 9+a>

+ 9 0

+ 2 +5o+1

+$9 +~z.

+q2 ~iso

+ 6

+$6-,~

+ 6

+ 7

+

-37

+ 0

+2 -v+0

~ ~5 105 127 129 114

~ 5<

121 126 N.A.

~7o HA

+ g-(

N.A.

H.A.

N.A.

H.A.

N.A.

(a)

= Determined per applicable ASME-BPY-Code Sect. III, Subsection HB, Article NB-2331-(a-1,2,3).

H.A. = Hot Applicable (no minimum upper shelf requirement).

= Lower bound curve values of transverse specimens.

0 l

Proposed Change and Reason for Change to Technical Specification Table 4.4-5 The proposed changes are on the attached marked up page.

Add to the LHAD FACTOR valuesc. 1.5, this is to be consistent with the PVNGS FSAR Section 5.3, Table 5.3-4A, sheets 1 through 6, which lists the Lead Factor as<.1.5.

0

TABLE 4.4-5 REACTOR VESSEL MATERIAL SURVEILLANCE PROGRAM " WITHDRAWAL SCHEDULE CAPSULE NUMBER VESSEL LOCATION 38 43 137 1424 230 310 LEAD FACTOR

< 1.5

+ 1.5 Q 1.5 g 1.5 Q 1.5

+1. 5 WITHDRAWAL TIME EFPY Standby Standby 4-6

-Standby 12 " 15 18 " 24 CS I

~ II

~

OI ~ lfP

~

$ II

Proposed Changes and Reason for Change to Technical Specifications 3/4.7.11 and 3/4.8.1.1 The proposed changes are on the attached marked up pages.

The basis for these changes is as follows.

The changes for both of the specifications are identical and involve the ASTH standards by which diesel fuel is sampled and analyzed.

Me are requesting that both of these specifications reflect the change that was granted to a special request for the Unit 1

Technical'pecifications prior to the full power license.

PLANT SYSTEMS SURVEILLANCE RE UIREMENTS (Continued) d.

e.

At least once per 6 months by performance of a system flush.

At least once per 12 months by cycling each testable valve in the flow path through at least one complete cycle of full travel.

f:

At least once per 18 months by performing a system functional test which includes simulated automatic actuation of the system throughout its operating

sequence, and:

l.

Verifying that each pump develops at least 1350 gpm at an indicated differential pressure of 125 psid by recording readings for at least 3 points on the test curve, 2.

Cycling each valve-in the flow path that is not testable during

~lant operation through at least one complete cycle of full

travel, and 3.

Verifying that each fire suppression pump starts sequentially to maintain the fire suppression water system pressure greater than or equal to 85 ps.ig.

At least once per 3 years by performing a flow test of the system in accordance with Chapter 5, Section ll of the Fire Protection

Handbook, 14th Edition, published by the National Fire, Protection Association.

4.7.11.1.2 The fire pump diesel engines shall be demonstrated OPERABLE:

a.

At least once per 31 days on a STAGGEREO TEST BASIS by verifying:

b.

1.

The diesel fuel oil day storage tanks each contain at least 315 gallons of fuel, and r

2.

The diesel engines start from ambient conditions and operate for at least 30 minutes on recirculation flow.

At least once per 92 days by verifying that a sample of diesel fuel from the fuel storage tank, obtained in accordance with ASTM-D~~gc6'p-8P is within'he acceptable limits specified in Table 1 of ASTM D975-~6'/

when checked for viscosity, water, and sediment.

c.

At least once per 18 months during shutdown, by subjecting the diesels to an inspection in accordance with procedures prepared in conjunction with its manufacturer's recommendations for the class of service.

, 4'.7. 11.1.3 Each fire pump diesel starting 24-volt battery bank and charger shall be demonstrated OPERABLE:

PALO VERDE - UNIT 2 3/4 7-30 OCT>> e6

1 ELECTRICAL POWER SYSTEMS SURVEILLANCE RE UIREMENTS 4.8. 1.1. 1 Each of the above required physically independent circuits between the offsite transmission network and the onsite Class 1E distribution system shall be:

a.

Determined OPERABLE at least once per 7 days by verifying correct breaker alignment indicating power availability Demonstrated OPERABLE at least once per 18 months during shutdown by manually transferring the onsite Class lE power supply from the normal circuit to the alternate circuit.

4.8.1.1.2 Each diesel generator shall be demonstrated OPERABLE:

a.

In accordance with the frequency specified in Table 4.8-1 on a STAGGERED TEST BASIS by:

1.

Verifying the fuel level in the day tank.

2.

Verifying the fuel level in the fuel storage tank.

3.

Verifying the fuel transfer pump can be started and transfers fuel from the Storage system to the day tank.

b.

C.

4.

Verifying the diesel generator can start** and accelerate to synchronous speed (600 rpm) with generator voltage and frequency at 4160 e 420 volts and 60 4 1.2 Hz in less than or equal to 10 seconds.

Subsequently, verifying the generator is synchronized, gradually loaded"" to an indicated 5200-5400 kW""" and operates for at least 60 minutes.

5.

Verifying the diesel generator is aligned to provide standby power to the associated emergency busses.

At least once per 92 days by verifying that a sample of diesel fuel from the fuel storage tank obtained in accordance with ASTM~~;Wc6'7"~l is within the acceptable limits specified in Table 1 of ASTM D975-~Sf when checked for viscosity, water and sediment.

I At least once per 184 days the diesel generator shall be started""

and accelerated to at least 600 rpm in less than or equal to 10 seconds.

The generator voltage and frequency shall be 4160 a 420 volts and 60 a 1.2 Hz within 10 seconds after the start signal.

)

""This test shall be conducted in accordance with the manufacturer's recommen-dations regarding engine prelube and warmup procedures, and as applicable regarding loading recommendations.

>""This band is meant as guidance to avoid routine over loading of the engine.

Loads in excess of this band for special testing under direct monitoring of the manufacturer or momentary variations due to changing bus loads shall not invalidate the test.

PALO VERDE -. UNIT 2 3/4 8-3

0 0

0

Proposed Change and Basis Change to Technical Specification 3.6.4.2 The proposed change is on the attached marked up pages.

The basis for the proposed change is as follows.

The Technical Specifications as presently written allow operation in iMODES 1 and 2 for an unlimited time with one hydrogen recombiner system INOPERABLE as long as the hydrogen purge system is OPERABLE.

The bases for this specification state that the recombiner or the purge unit are capable of controlling the expected hydrogen generation during an accident.

Ne believe that since there is a separate specification for the purge unit, and that this unit is capable of controlling the expected hydrogen generation during an accident, the proposed change is prudent and does not increase the risk to the public.

O.

i 0

CONTAINMENT SYSTEMS ELECTRIC HYDROGEN RECOMBINERS LIMITING CONDITION FOR OPERATION 3.6.4.2 Two portable independent containment hydrogen recombiner systems shared among the three units shall be OPERABLE.

APPLICABILITY:

MODES 1 and 2.

ACTION:

With one hydrogen recombiner system inoperable, restore the inoperab'le system to OPERABLE status within 30 days or meet the requirements of Specification 3. 6.4; 3, or be in at least HOT STANDBY within the next 6 hour6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months s.

~~yp ~~~ gycq~spl REcowsIhrBg Gysren uvopFRAslE, AHD re~ epo~oc s> ~>w~<

sgsf'EA opaAA84K P<g S PEcxcs'M7zoA g,6. g. )

vga'govlslunrs a+ SPY i ~c4ncw

$ o.y P4F A nr SPPaiCASaE, SURVEILLANCE RE UIREMENTS 4.6.4.2 Each hydrogen recombiner system shall be demonstrated OPERABLE:

a.

At least once per 6 months by:

1.

Verifying through a visual examination that there is no evi-dence of abnormal conditions within the recombiner enclosure.

and control console.

2.

Operating the air blast heat exchanger fan motor and encIosed blower motor continuously for at least 30 minutes.

b.

At least once per year by:

1.

Performing a

CHANNEL CALIBRATION of recombiner instrumentation.

2.

Performing a "Low-Level Test-Heater Power Off" and "Low-Level Test-Heater Power On" test and verifying that the recombiner temperature increases to and is maintained at 600 a 25'F for at least one hour.

With power off and a simulated input signal of 1280'F, verify the OPERABILITY of all control circuits.

When this test is conducted, the air blast heat exchanger fan motor and enclosed blower motor shall be'perated continuously for

. at least'0 minutes.

,C.

At least once per 5 years by performing a Recombiner System High-Level Test" and verifying that the recombiner temperature%

increases to and is maintained at 1200 2 50 F for at least one hour.

PALO VERDE - UNIT 2 3/4 6-37 OCT 1l 88'

0 0

Proposed Changes and Basis for Changes to Technical Specification 3.6.4.1 and Bases Located on Page B3/4 6-4 The proposed changes are on the attached marked up pages.

The basis for the changes is as follows:

The containment a'tmosphere hydrogen and oxygen monitors have been deleted from the design of the PASS.

Region V has accepted our current design and plans to obtain and anlyze grab samples to meet the requirements of NUREG 0737 Section II.B.3 and Regulatory Guide 1.97.

CONTAINMENT SYSTEMS

'3/4.6.4 COMBUSTIBLE GAS CONTROL HYDROGEN MONITORS LIMITING CONDITION FOR OPERATION 3.6.4. 1 Two independent containment hydrogen monitors shall be OPERABLE.

APPLICABILITY:

MODES 1 and 2.

ACTION:

a.,

With one hydrogen monitor inoperable, restore the inoperable monitor to OPERABLE status within 30 days or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ; b.

With both hydrogen monitors inoperable, restore at least one monitor to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

~Pg~A g C4,PABEITP'llh hy l>>,

h h

ygthe Post Accident Sampling System OPERABLE, the provisions of Specification 3.0.4 are not applicable.

SURVEILLANCE RE UIREMENTS

4,.6.4.1 Each hydrogen monitor shall be demonstrated OPERABLE by the performance of a CHANNEL CHECK at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , a

CHANNEL FUNCTIONAL TEST at least once per 31 days, and at least once per 92 days on a STAGGERED TEST BASIS by performing a CHANNEL CALIBRATION using sample gases containing a nominal:

a..

One volume percent

hydrogen, balance nitrogen.

b.

Four volume percent

hydrogen, balance nitrogen.

. PALO VERDE - UNIT 2 3/4 6-36 DCT 11 mS

~

CONTAINMENT SYSTEMS BASES 3/4. 6. 3 CONTAINMENT ISOLATION VALVES The OPERABILITY of the containment automatic isolation valves ensures that the containment atmosphere will be isolated from the outside environment in the event of a release of radioactive material to the containment atmosphere or pressurization of the containment and is consistent with the requirements of GDC 54 through GDC 57 of Appendix A to 10 CFR Part 50:

Containment isolation within the time limits specified for those isolation valves designed to close automatically ensures that the release of radioactive material to the environ-ment will, be consistent with the assumptions used jn the analyses for a LOCA.

The only valves in the Table 6.2.4-1 of the PVNGS FSAR that are no't required to be listed in Table 3.6-1 are the following:

main steam safety valves and main steam atmospheric dump valves.

The main steam safety valves

'nd the atmospheric dump valves have very high pressure setpoints to actuate and are covered by Specifications 3/4. 7. 1.1 and 3/4. 7. l. 6, respectively.

3/4.6.4 COMBUSTIBLE GAS CONTROL The OPERABILITY of the equipment and systems required for the detection and control of hydrogen gas ensures that this equipment will be available to maintain the hydrogen concentration within containment below,its flammable limit during post-LOCA conditions.

Either recombiner unit (or the purge system) is capable of controlling the expected hydrogen generation associated with (1) zirconium-water reactions, (2) radiolytic decomposition of water and (3) corrosion of metals within containment.

These hydrogen control systems are consistent with the recommendations of Regulatory Guide 1.7, "Control of

" Combustible Gas Concentrations in Containment Following a LOCA," March 1971.

There is a hydrogen monitor and a oxygen monitor in the Post Accident Sam-pling System (PASS).

There is a line from the two containment hydrogen monitors to the PASS which allows the licensee to monitor hydrogen concentration in the containment following an accident with the PASS.

An OPERABLE hydrogen monitor in the PASS allows the licensee to enter Modes 1 and 2 with one containment hydrogen monitor inoperable.

For the hydrogen monitor in PASS to be

OPERABLE, the valves in the piping to the monitor must be OPERABLE.

gap~Aca'Hs<~

PA<Ac~RePg There is a line from one of the containment hydrogen monitors to the PASS which allows for sampling of post accident hydrogen concentration in the containment.

Sampling of the containment atmosphere in this manner meets

. the requirements of NUREG 0737 Section II.B.3 and Regulatory. Guide 1.97.

An OPERABLE sampling capability in the PASS allows the licensee to enter MODES l and 2 with one containment hydrogen monitor inoperable.

For the hydrogen sampling capability to be OPERABLE the valves in the piping to the sample point must be operable.

PALO VERDE " UNIT 2 B 3/4 6-4 Qggi W

Proposed Changes and Basis for Changes to Technical Specification 3/4.3.1 The proposed changes are on the attached marked up pages.

The bases for these changes is as follows.

1.

Page 3/4 3-3 deletion of 3 and 4 to applicable modes:

MODE 3 and 4 were added because of NRC proposed changes to the Standard Technical Specifications for CE plants.

CE does not have not have these changes in their copy of these Specifications.

In addition, no credit is taken for the low pressurizer pressure.trip with the plant in MODE 3 and 4 in the Safety Analysis.

We believe that the addition of MODE 3 and 4 to the applicable modes does not add a margin of safety.

2.

Page 3/4 3-14, 15 and 16 These changes restore a footnote to the table which probably should not have been deleted.

The table notation and footnotes are being numbered as they appear in'he Unit 1 Tech'nical Specifications.

The table n'ota-tion is being changed to reflect the deletion of Specification 2.2.2.

t

TABLE 3.3-1 REACTOR PROTECTIVE INSTRUMENTATION FUNCTIONAL UNIT I.

TRIP GENERATION TOTAL NO.

OF CHANNELS MINIMUM CHANNELS CHANNELS APPLICABLE TO TRIP OPERABLE MODES ACTION A.

Process 1.

Pressurizer Pressure " High 2.

Pressurizer Pressure Low 3.

Steam Generator Level - Low 4/SG 4.

Steam Generator Level - High 4/SG 5.

Steam Generator Pressure

- Low 4/SG

6. "Containment Pressure - High 4

7.

Reactor Coolant Flow - Low 4/SG 8.

Local Power Density - High 4

9.

DNBR " Low B.

Excore Neutron Flux 2

2 (b) 2/SG 2/SG 2/SG 2

2/SG 2 (c)(d) 2 (c)(d) 3 3

3/SG 3/SG 3/SG 3

3/SG 3

3 "Cb 1 $ 2 1,

2 1, 2, 3*, 4*

1 2

1>>

2 1

2 2¹ 3¹ 2¹ 3¹ 2¹ 3¹ 3¹ 2¹ 3¹ 2¹ 2¹, 3¹ 2.

Variable Overpower Trip Logarithmic Power Level - High a.

Startup and Operating

~ b.

Shutdown I

C.

%re )rotection Calculator System 1.

CEA Calculators 2.

Core Protection Calculators 4

1 2 (c)(d) 2 (e) 3 2 (a)(d) 2 3

0 2

1, 2

3A 4A 5A' 3, 4, 5

1, 2

2~

¹ 'm 8

p'=:m

~r4rE'.>

<<c3 6,

7 2¹ 3¹a ~

0

TABLE 4.3-1 REACTOR PROTECTIVE INSTRUMENTATION SURVEILLANCE RE UIREMENTS FUNCTIONAL UNIT I.

TRIP GENERATION A.

Process 1.

Pressurizer Pressure - High 2.

Pressurizer Pressure - Low 3.

Steam Generator Level - Low 4.

Steam Generator Level - High 5." Steam Generator Pressure - Low 6.

Containment Pressure

- High 7.

Reactor Coolant Flow - Low 8.

Local Power Density " High 9.

DNBR - Low B.

Excore Neutron Flux CHANNEL CHANNEL CHECK CALIBRATION R

R R

1 R

R D (2, 4),

R (4, 5)

D (2, 4),

R (4, 5)

M (8}

S (7)

CHANNEL FUNCTIONAL TEST

.M M

M M

M M,

R (6)

M, R (6)

MODES IN MHICH SURVEILLANCE RE UIRED 1

2 1,

2 1

2 3*

4*

1) 2 1,

2 1

2 1, 2, 1.

Variab'1e Overpower Trip I

2.

Logarithmic Power Level - High C.

'Co're Protection Calculator System

~ i +o.nr

~l 1:

CEA Cal cul ators 2.

Core Protection Calculators 0 (2, 4),

M (3; 4) 0 (4)

R (4)

R D (2, 4),

R (4, 5)

M (8),

S (7) 1 p 2

M,'R (6)

'1, 2

R (6)'"'

1 2

(~),

wg r.5 s~i3 "Vg M and S/U (1) 1, 2, 3, 4, 5

and "

RPjf

~ie Cg g~

t

TABLE 4.3-1 (Continued)

REACTOR PROTECTIVE INSTRUMENTATION SURVEILLANCE RE UIREMENTS M

FUNCTIONAL UNIT CHANNEL CHANNEL CHECK CALIBRATION CHANNEL FUNCTIONAL TEST MODES IN WHICH SURVEILLANCE REIEI ERR D.

Supplementary Protection System Pressurizer Pressure - High 1,

2 II.

RPS LOGIC A.

Matrix Logic N.A.

N.A.

3A 4A 5A'.

Initiation Logic N.A.

N. A.

1, 2, 3~, 4*, 5" III. RPS ACTUATION DEVICES A.

Reactor Trip Breakers B.

Manual Tr ip N.A.

N.A.

N, R~gio) 1, 2, 3", 4", 5' 3A 4A 5*

~

%1 ~

E4 t I

~

I Cln

TABLE 4.3-1 (Continued)

F)fhj Ps I

gw p p'7 TABLE NOTATIONS (2)

(4)

(5)

(6)

(7)

(8) 5Q

, le.

OCT 11 1985 3/4 3-16 PALO VERDE -. UNIT 1 With reactor trip breakers in the closed position and the CEA drive system capable of CEA withdrawal, and fuel in the reactor vessel.

Each STARTUP or when required with the reactor trip breakers closed and the CEA drive system capable of rod withdrawal, if not performed in the previous 7 days.

Heat balance only (CHANNEL.FUNCTIONAL TEST not included),

above 15K of RATED THERMAL POWER; adjust the linear power level, the CPC delta T

power and CPC nuclear power signals to agree with the calorimetric calculation if absolute difference is greater than 2X.

During PHYSICS TESTS, these daily calibrations may be suspended provided these calibrations are performed upon reaching each major test power plateau and prior to proceeding to the next major test power plateau.

Above 15K of RATED THERMAL POWER, verify that the linear power sub-channel gains of the excore detectors are consistent with the values used to establish the shape annealing matrix elements in the Core

, Protection Calculators.

Neutron detectors may be excluded from CHANNEL CALIBRATION, After each fuel loading and prior to exceeding

?OX of RATED THERMAL POWER, the incore detectors shall be used to determine the shape annealing matrix elements and the Core Protection Calculators shall use these elements.

This CHANNEL FUNCTIONAL TEST shall include the injection of simulated process signals into the channel as close to the sensors as practicable to verify OPERABILITY including alarm and/or trip functions.

Above

?OX of RATED THERMAL POWER, verify that the total steady-state RCS flow rate a's indicated by each CPC is less than or equal to the actual RCS total flow rate determined by either using the reactor coolant pump differential pressure instrumentation or by. calorimetric calculations and if necessary, adjust the CPC addressable constant flow coefficients such that each CPC indicated flow is less than or equal to the actual flow rate.

The flow measurement uncertainty may be included in the BERRl term in the CPC and is equal to or greater than 4X.

Above ?OX of RATED THERMAL POWER, verify that the total steady-state RCS flow, rate as indicated by each CPC is less than or equal to the actual RCS total flow rate determined by either 'using the reactor coolant pump differential pressure instrumentation and the ultrasonic flow meter adjusted pump curves or calorimetric calculations.

At least once per 18 months and following maintenance oZ adjustment of the reactor trip breakers, the CHANNEL FUNCTIONAL TEST shall include independent ver'ification of the undervoltage anil shunt trips.

P4

'Tke. movh44y c.RAHU~<~c~~~u~se zMLw~ga I'<p~ <<a ego~ (km'W c.asra c 4 (c~r~h) aralu ~ aj 4~~4 ~~~.~%~a.~h.~ c.~A.

QPPRAQ~G

<PC.

t I

Proposed Change and Basis for Change to Technical Specification 3.6.1,.3 The proposed change is on the attached marked up'age.

Basis for the change is as follows.

Allowing the outer door to be opened for a small fraction of the time during the year will allow maintenance on the inner door of the airlock.

Restricting the amount of time the outer door is open, insures that the probability of risk to the health and safety of the public remains low.

This change is similar to the exception being granted to Millstone Point 3.

CONTAINMENT SYSTEMS CONTAINMENT AIR LOCKS LIMITING CONDITION FOR OPERATION I

tidal t~g 3.6. 1.3 Each containment air lock shall be OPERABLE with:

a.

Both doors closed except when the air lock is being used for normal transit entry and exit through the containment, then at least one air lock door shall be closed, and 4

b.

An overall air lock leakage rate of less than or equal to 0.05 L

at P,

49.'Z psig.

-5 APPLICABILITY:

MODES 1, 2, 3, and 4.

ACTION:

a.

b.

With one containment air lock door inoperable:

1.

Maintain at least the OPERABLE air lock door closed and either restore the inoperable air lock door to OPERABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or lock. the OPERABLE air lock door closed.

Operation may then continue until performance of the next required overall air lock leakage test provided that the OPERABLE air lock door is verified to be locked closed at least once per 31 days, or 2.

Be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

3.

The provisions of Specification 3.0.4 are not applicable.

With the containment air lock inoperable, except as the result of an inoperable air lock door, maintain at least one air lock door closed; restore the inoperable air lock to OPERABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE RE UIREMENTS 4.6. 1.3 Each containment air lock shall be demonstrated OPERABLE:

a.

Within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months following each closing, except when thewir lock is being used for multiple entries, then at lea'st once per p2 hours, by verifying seal leakage to be less than or equal to 0.01 d.'hen determined with the volume between the door seals'pressurized to greater than or equal to 14.5 s 0.5 psig, for at least -15 minutes, g Fxcerr ouaiIJG- 'Evvray pn pspplg pd

)hluE'8, boa/> AA, Cu~qt.Arly 7VmF uoP ~Ex'cEFD

( Hood

~~iC /~<4.

PALO VERDE - UNIT 2 3/4 6"4 CP2) ~g

Basis For Change To Technical Specification 3/4.2.5 PVNGS Unit 1 Technical Specification 3/4.2.5 (RCS Flowrate) required a minimum actual flowrate of 164.0x10 ibm/hr.

Flowrate was measured during post-core hot functional testing with an ultrasonic flow meter (UFM).

The measured flowrate at zero power was 105.1% of the design flowrate.

This value must be adjusted by two factors before it can be compared with the technical specifi-cation flow requirement.

The first factor accounts for the fact that the primary coolant specific volume on the core exit side is higher during full power operation than it is at zero power.

The currently projected value associated with the resulting flowrate reduction is 1.4X

Thus, the projected full power flowrate is 103.7X.

The second factor accounts for the uncertainty associated with the flow measurement.

The attached figure shows the flowrate measurement uncertainty associated with the mass flowrate indicated by COLSS, designated as NKTMFLOW.

The variation in the flow measurement uncertainty with time is the result of drift and'adiation effects on the pump differential pressure transducer output.

As can be seen from the attached figure, the uncertainty associated with 4 months of operation is 3.4X.

Using this value for illustration pur-

poses, the projected full power flowrate adjusted for measurement uncertainty is 100.3X.

N The previously mentioned technical specification as presently written is unclear as to the meaning of actual flowrate which is to be compared to 164 x 10 ibm/hr (100X design flow).

Comparing indicated flowrate to 100X is bounded by the safety analysis.

C-E's original interpretation of this technical specification however required subtraction of the flowrate measurement uncertainty from the indicated flowrate before comparison with the 100X value.

Using a projected indicated full power flowrate of 103.7X and an uncertainty of.-

103.7X and an uncertainty of 3.4X, this comparison results in a +0.3X "operational flowrate margin" at 4 months from initial calibration of the pump differential pressure transducers.

At the end of cycle 1 the operational flowrate margin for PVNGS Unit 1 could be near zero.

The operational flowrate margin may also be diminished by other factors.

These are listed below.

1.

Core crudding:

1 psi of core crudding causes a flowrate reduction of 0.35X.

2.

Tube plugging: Plugging 100 short steam generator tubes causes an estimated flowrate reduction of 0.18X.

3.

Random effects associated with each flow measurement can influence a given measurement in the positive or negative direction.

According to the corresponding basis section, the flowrate requirement is provided to ensure that the safety analysis assumptions are maintained.

In the safety analyses, the effects of flowrate are evaluated over a range of 95 116X Where flowrate impacted the anlysis results, the most adverse flow-rate was used.

Where flowrate )id not significantly affect analysis results a nominal value of 100X (164x10 1bm/hr) was used in conjunction with a sensitivity analysis.

Thus, we believe that6the safety analysis assumptions are preserved by a flow rate of 155.8xl0 ibm/hr (95X of 164x10 ibm/hr).

Since the operational flowrate margin is small, we are requesting that the specified value for RCS flowrate be changed to the 95/. value.

0 0

0 II

~

s sa ~ 5 ses Ssa ssta st; ss ceases

~.

s Sstte

~

a s

= ~ e

~ ~ av

~ ~ aiae.

~

46 1323 Attachment to

-32599 I

s s

~

'I s

~

I I

s I

I I

~

I I

I

'f I

I' puCCPTP <<MTY I

ELHI uL'0 I

~

s I

I I

e I

~ ~

PU'~l H.C!

I

,s)i ls.

I ~

tse

'I s

AC(VTA<<QF/ l

)e!A

.i.

p oF,DQ<<6d I r)A'Z 'P'LOsJ I

i R<t -

l

~ l I

s I

. ~

I elI' I'

a I

I e

s I

i

~

~ il s

Is tss I

I I

a

~~~ ~

~

~II~

I s

l

is I

e s

I f

~

I I

~ ~

,I

~ '

I I

PPLtsaPS)1)

I I

s I

~

l

,'ai

I a..

s I

~

st I

)

I

~

s ~ '

e I

~

I si

, lit sl 's

~ i

~

I

~ ~

I ~

e I

lt ~ I I

s s

~ 's I

e

~"I e

sii. ~

s

~ ie

)

I PHE' s

I l

l s

I "i'.:

II.

s-i~

H SUR~:

I I'ness 7.

s s'

~

sit si

~

I

's,i

~ I

,.'i 1

I!".I s.

I ~

l., i

~ i>>

I I;

\\

~

I si is s

I s)

L s

I e

~

I

I'

~

Int

, I'

'I s e.s.a s

) s pi I

~ ~

I

~ '

I;'

I

il' s

~

I s

tl ~

I I

i s IIi i'is

~

~ e i'!

<<I

~ )I

"'I

'-'I ~

I':

I'I'li

'tl,

I

-I!i

)le

~.e)'

~ ~s, e

)

s,I

~

I j<<

~

li I I '. ~

s s l" i<<j

~

s

~

e s

s e

)

s it:'I B A s

s s

~

i.'..

':I Is;;

s

',:ll s

,)

s'is

~ ~

is<<

el<<

il f

L l

,lsl II::

i ~

'itf i'st j

I.f I

0 I'

i e

~ I e

~

s

~ :I~

~,Is j ssl

".ss'

~ 'Is s

st

~

~ ~ ~

I I I

I as s

I.

e

I'

'i I s 'I I':l l

s

'l s e li

~

e it's

~

s

~ s I s'. sl s

s I'

s:I s

sI'l!

I '.

s s

~ s s

-: I

)st I'I:

~

'iI)

1

~ i]!

sl:,

lt I

I I.

) I;.:

I el)ts P

st I

'I't<<;

I s'

ef:

-I i)

I,'r:l s

~

Is s

s

~

~ I I

~ )

~,tl s

I

I:

I) e I',I:

'sf s

~ I

~ ~ i s

~

ffI

'it

~.II jf; ml:

I ~

s i..t I'I ii<<;

ifl:

~

s

- og SL'UC.

~

~ s I

I

~ ls-s I"-

li-ll-il

~ ~

ll e

,'I'

-I a

st st s

s I

I I

s e

S a

s I.'.

I::l s

t'

e. ~

e I 'I s

~

e

~

I ~ ~

s e

~

s ~

I

~ Se

I e

. ~

le a

I s.J..

ATE'hl s

s

's I

~

I

...I

)'Z

)

ilj s

~

s

.)I s

s s

sa e

s I

s '

1

';ll.:

j'r

i:!

II!-'ll<<

l:I Tl

,s UC LIu3

-ts s

T.

l,i;

i-'t ri st I

I PT!

I!::,,

lt

~

)'I

~ it I:.

I'.I,

]I)t il jIli

)e)s j'-'ss'

~

I

'I I

",I s

~ '

s

';]l lijt

~ s

':I I

.)il slit

.I, lil!

.I,f

~ '

Ill if s

s

~

If/<<

.l';sf e

qJ I i

~ e

~

s s

I

~ a r

I:

~ Il

~

,l'l

~

s s

I'

~

is)I

~ I si e

<<'II s

~

I ~

)I ~

frI I

,"I:

s is

~.I

~ sl ~

"-"Il I

s

'I r.,

~ sest s

-Ho I "'

~ ~

.j...

s'al>>

a s

e

~

,s

~

s

~

s i

~ I I

~ s jtli

'.I j s)if

'.el.

I I

lml; Iij,l'/

i jli Illi I:i'

~ i (

It

!II I

I

I:I i<<i, lit',

III I

s l;

s I

~

~ f l

s

~

'rjj I',l:

I' s

s

~

aitiI:

tie ii't I

~ ~

);..

slt

li sI)l I

i

~

si' I

'<<I..

. I.

I

~ st Iili l'lls

'.I!.;

s

~ ~

~lf,'

~ s Il

.ill

!it:

s j)lt

I'.,

lf fll' I'l, lss

I!:;

I.e Is e I e ~

0 0

0

POWER DISTRIBUTION LIMITS 3/4. 2. 5 RCS FLOW RATE LIMITING CONDITION FOR OPERATION 3.2.5 The a'ctual Reactor Coolant System total flow rate shall be greater than or equal to ~:Kx 10e ibm/hr.

>55.

APPLICABILITY:

MODE l.

ACTION:

With the actual Reactor Coolant System total flow rate determined to be less than the above limit, reduce THERMAL POWER to less than 5X of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

SURVEILLANCE RE UIREMENTS 4.2.5 The actual Reactor Coolant System total flow rate shall be determined to be greater than or equal to its limit at least'nce per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

4

~

C 4

PALO VERDE - UNIT 2 3/4 2-e OCT 11 1%5

0

Proposed Changes and Basis for Changes to Technical Specification 3/4.6.1 The proposed changes are on the attached marked up pages.

Pages 3/4 6-1 6-2 6-3 6-4 6-5 6-15 B 3/4 6-2 The bases for these changes are enclosed; ANPP letter 33610-EEVB/KLH, dated

.September 30, 1985.

0

3/4. 6 CONTAINMENT SYSTEMS 3/4. 6. 1 PRIMARY CONTAINMENT CONTAINMENT INTEGRITY LIMITING CONDITION FOR OPERATION 3.6. 1.1 Primary CONTAINMENT INTEGRITY shall be maintained.

APPLICABILITY:

MODES 1, 2, 3, and 4.

ACTION:

Without primary CONTAINMENT INTEGRITY, restore CONTAINMENT INTEGRITY within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE RE UIREHENTS

4. 6.1.1 Primary CONTAINMENT INTEGRITY shall be demonstrated:

a 0 b".

C.

At least once per 31 days by verifying that all penetrations" not capable of being closed by OPERABLE containment automatic isolation valves and required to be closed during accident conditions are closed by valves, blind flanges, or deactivated automatic valves secured in their positions except as provided in Table 3.6-1 of Specification 3.6.3.

By verifying that each containment air lock is in compliance with the requirements of Specification 3.6. 1.3.

After each closing of each penetration subject to Type B testing-,

except containment air locks, if opened following a Type A o 8

test, by leak rate testing the seal with gas at P

49.+ sig an verifying that when the measured leakage rate for these seals is added to the leakage rates determined pursuant to Specifica-tion 4.6.1.2d. for all other Type B and C penetrations, the combined leakage rate is less than or equal to 0.60 L

a'xcept valves, blind flanges, and deactivated automatic valves which are located inside, the containment and are locked, sealed, or otherwise secured in the closed position. 'hese penetrations shall be verified closed during each COLD SHUTDOWN except that such verification need not be performed more often than once per 92 days.

PALO VERDE " UNIT, 2 3/4 6-1 CCT 1] mz

CONTAINMENT SYSTEMS CONTAINMENT LEAKAGE FÃ~V~II PRPg LIMITING CONDITION FOR OPERATION 3.6.1.2 Containment leakage rates shall be limited to:

a.

An overall integrated leakage rate of:

1.

Less than or equal to L

, OelOX by weight o e

containment air per 24 iIoore at P

, 49.$

eig, or a'.

Less than or equal to L

, 0.05K by weight of the containment air per 24 (ours at a reduced pressure of Pt, 24.6 psig.

b.

A combined leakage rate of less than or equal to 0.60 L

for all penetrations and valves subject to Type B and C tests, Shen pressurized to P

1 APPLICABILITY:

MODES 1, 2, 3, and 4.

ACTION:

Mith either (a) the measured overall integrated containment leakage rate exceeding 0.75 L

or 0.75 L, as applicable, or (b) with the measured combined leakage rate for,all penetrktions and valves subject to Types B and C tests exceeding

0. 60 L

, restore the ovei all integrated leakage rate to less than or equal to 0. 75 L

or less than or equal to 0. ?5 L

as applicable, and the combined leakage rate for all penetrations and valves subject to Type B and C

tests to less than or equal to 0.60 L prior to increasing the Reactor Coolant System temperature above 210 F.

SURVEILLANCE RE UIREMENTS 4.6.1.2 The containment leakage rates shall be demonstrated at the following test schedule and shall be determined in conformance with the criteria specified in Appendix J of 10 CFR Part 50 using the methods and provisions of ANSI N45.4 - 1972:

a.

Thiee Type A tests (Overall Integrated Containment Leakage Rate) shal.l be conducted at 40 + 10 month intervals during shutdown at either P

49.

psig or at P

24.6 psig during each 3.0-year service period.

The hird test of $ach set shall be conducted during the shutdown for the 10-year'plant inservice inspection.

m I

PALO VERDE - UNIT. 2 3/4 6-2 OCT)I 8 ~

CONTAINMENT SYSTEMS

,~~3.g,'ij s)'URVEII LANCE RE UIREMENTS Continued b.

-,If any periodic Type A test fails to meet either, 0.75 L-or 0-75 L, the test schedule for subsequent Type A tests shall be Reviewed an) approved by the Commission.

If two consecutive Type A tests fail to meet either 0.75 L

or 0.75 L

a Type A test shall be performed at least every 18 months until two consecutive Type A tests meet either 0.75 L

or 0.75 L

at which time the above test schedule may be resumei.

c.

The accuracy of each Type A test shall be verified by a supplemental test which:

d.

1.

Confirms the accuracy of the Type A test by verifying that the supplemental test result L

minus the sum of the Type A test

result, L

, and the superimposed leak rate, L, is equal to or less'tMn 0.25 L.

a 2.

Has a duration sufficient to establish accurately the change in leakage rate between the Type A test and the supplemental test.

3.

Re uires that t e ra e at w i l.

Air locks, q

h t

h ch gas ss sngected onto the contain-ment or bled from the containment during the supplemental test is between 0.75 L

and 1.25 L.

a 5'Type B and C tests shall be conducted with gas at P

, 49.$ psig, at intervals no greater than 24 months except for tests involvirig:

2.

Purge supply and exhaust isolation valves with resilient material seals.

e.

Purge supply and exhaust isolation valves with resilient material

'-seals-shall be tested and demonstrated OPERABLE per Specifications 4.6.1.7.2 and 4.6.1.7.3.*

Air locks shall be tested and demonstrated OPERABLE per Specification

4. 6. 1. 3.

g.

The prov'isions of Specification 4.0..2 are not applicable.

PALO VERDE - UNIT 2 3/4 6"3 OCTgg

CONTAINMENT SYSTEMS CONTAINMENT AIR LOCKS LIMITING CONDITION FOR OPERATION 3.6.1.3 Each containment air lock shall be OPERABLE with:

a.

Both doors closed except when the air lock is being used for normal transit entry and exit through the containment, then at least one air lock door shall be closed, and b.

An overall air lock leakage rate of less than or equal to 0.05 L

at P, 49.+psig.

r Q APPLICABILITY:

MODES 1, 2, 3, and 4.

ACTION:

a.

With one containment air lock door inoperable:

1.

Maintain at least the OPERABLE air lock door closed and either restore the inoperable air lock door to OPERABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or lock.,the OPERABLE air lock door closed.

Operation may then continue until performance of the next required overall air lock leakage test provided that the OPERABLE air lock door is verified to be locked closed at least once per 31 days, or 2.

Be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

3.

The provisions of Specification

3. 0.4 are not applicable.

b.

With the containment air lock inoperable, except as the result of an inoperable air lock door, maintain at least one air lock door closed; restore the inoperable air lock to OPERABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE RE UIREHENTS 4.6.1.3

'Each containment air lock shall be demonstrated OPERABLE:

a.

Within.72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months following each closing, except when themir lock is being used for multiple entries, then at least once per g2 hours, by verifying seal leakage to be less than or equal to 0.013'hen 0

determined with the volume between the door seals pressurized to greater than or equal to 14.5 2 0.5 psig, for at least -15 minutes, PALO VERDE - UNIT-2 3/4 6-4 CCT 1I ~q

~,

CONTAINMENT SYSTEMS SURVEILLANCE RE UIREHENTS (Continued

))

'> -" (."s5 b.

C By conducting overall air lock leakage tests at not less than P

4S.

psig, and verifying the overall air lock leakage rate is'within its limit:

5 l.

At least

.once per 6 months,

and 2.

Prior to establishing CONTAINMENT INTEGRITY when maintenance has been performed on the air lock that could affect the air lock sealing capability~

C.

At least once per 6 months by verifying that only one door in each air lock can be opened at a time.

The prov>sions of Specification 4.0.2 are not applicable.

"This constitutes an exemption to Appendix J of 10 CFR Part 5Q.

PALO VERDE - UNIT 2 3/4 6-5

0

CONTAINMENT SYSTEMS 3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS CONTAINMENT SPRAY SYSTEM LIMITING CONDITION FOR OPERATION 3.6.2.1 Two independent containment spray systems shall be OPERABLE with each spray system capable of taking suction from the RWT on a containment spray actuation signal and automatically transferring suction to the containment sump on a recirculation actuation signals Each spray system flow path from the containment sump shall be via an OPERABLE shutdown cooling heat exchanger.

APPLICABILITY:

MODES 1, 2, 3, and 4,"..

gY With one containment spray system inoperable, restore the inoperable spray system to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT STANDBY'within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ; restore the inoperable spray system to OPERABLE status within the next 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or be in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE RE UIREMENTS 4.6.2.1 Each containment spray system shall be demonstrated OPERABLE:

b.

C.

d.

At least once per 31 days by verifying that each valve (manual, power-operated, or automatic) in the flow path is positioned to take suction from the RWT on a containment spray actuation (CSAS) test signal.

g5g By verifying that each pump develops anlindicated differential pressure of greater than or equal to hasid at greater than or equal the minimum allowable recirculati'on flowrate when tested pursuant to Specification 4.0.5.

At least once per 31 days by verifying that the system piping is full of water to the 60 inch level in the containment spray header

(>115 foot level).

At least once per 18 months, during shutdown, by:

l.

Verifying that each automatic valve in the flow path actuates to its correct position on a containment spray actuation (CSAS) and recirculation actuation (RAS) test signal.

2.

- Verifying that upon a recirculation actuation test signal, the containment sump isolation valves open and thaf a recirculation mode, flow path yia an OPERABLE shutdown cooling heat exchanger is established.

"Only when shutdown cooling is not in operation.

PALO VERDE - UNIT 2 I

3/4 6"15 OCT 11 ]g8g

CONTAINMENT SYSTEMS BASES 3/4.6.1.4 INTERNAL PRESSURE The limitations on containment internal pressure ensure that (1) the containment structure is prevented from exceeding its design negative pressure differential with respect to the outside atmosphere of 4 psig and (2) the containment peak pressure does not exceed the design pressure of 60 psig during LOCA conditions.

.5 The maximum peak pressure expected to be obtained from a LOCA event is

'49.$ psig.

The limit of 2.5 psig for initial positive containment pressure will limit the total pressure to 49.

psig which is less than the design pressure (60 psig) and is consistent with the safety analyses.

3/4. 6.l. 5 AIR TEMPERATURE The limitation on containment average air temperature ensures that the overall containment average air temperature does not exceed the initial tem-perature condition assumed in the safety analysis.

3/4. 6.l. 6 CONTAINMENT STRUCTURAL INTEGRITY i5 This limitation ensures that the structur

>integrity of the containment will be maintained comparable to the origina design standards for the life of the facility.

Structural integrity is requ ed to ensure that the containment will withstand the maximum pressure of 49.+ psig in the event of a LOCA.

The containment design pressure is 60 psig.

The measurement of containment tendon lift"offforce; the tensile tests of the tendon wires or strands; the examina-tion and testing of the sheathing filler grease; and the visual examination of tendon anchorage assembly

hardware, surrounding concrete and the exterior sur-faces of the containment are sufficient to demonstrate this capability.

The tendon wire or strand samples will also be subjected to tests.

All of the required testing and visual examinations should be performed in a time frame that permits a comparison of the results for the same operating history.

~ The Surveillance Requirements for demonstrating the containment's, structural integrity are in compliance with the recommendations of Regulatory Guide 1.35, "Inservice Surveillance of Ungrouted Tendons in Prestressed Concrete Containment Structures,'-'evision i 1984,;

~

II

'L9"rq The required Special Reports from any engineering evaluation of containment abnormalities shall include a description of the tendon condition, the condition of the concrete (especially at tendon anchorages),

the inspection procedures, the tolerances on cracking, the results of the engineering evaluation, and the corrective actions taken.

PALO VERDE - UNIT 2 B 3/4 6-2 OCT Sl $8~

U'

Proposed Changes and Basis for Changes To Technical Specification 3/4.6.1.6 The proposed changes are on the attached marked-up page.

The basis for the change is as follows:

Regulatory Guide 1.35, "Inservice Surveillance of Ungrouted Tendons in Prestressed Concrete Containment Structures",

Revision 3, 1984, should be changed to Revision 1,

1974.

In section 1.8 of the PVNGS

FSAR, we committed to Revision 1, 1974; except for the positions as stated in the FSAR.

Regulatory Guide 1.35.1, "Determining Prestressing Forces for Inspection of Prestressed Concrete Containments",

1984.

This Regulatory Guide has not been issued by the NRC, therefore, to reference to this document is incorrect.

0

CONTAINMENT SYSTEMS V QJ t BASES 3/4.6. 1.4 INTERNAL PRESSURE The limitations on containment internal pressure ensure that (1) the containment structure is prevented from exceeding its design negative pressure differential with respect'o the outsjde atmosphere of 4 psig and (2) the containment peak pressure does not exceed the design pressure of 60 psig during LOCA conditions.

,P The maximum peak pressure expC.cted to be obtained from a LOCA event is 49.g psig.

The limit of 2.5 psig for initial positive containment pressure will limit the total pressure to 49.+ psig which is less than the design pressure (60 psig) and is consistent with)the safety analyses.

3/4. 6. l. 5 AIR TEMPERATURE The limitation on containment average air temperature ensures that the overall containme'nt average air temperature does not exceed the initial tem-perature condition assumed in the safety analysis.

0 3/4. 6. 1. 6 CONTAINMENT STRUCTURAL INTEGRITY This limitation ensures that the structural integrity of the containment will be maintained comparable to the original design standards for the life of the facility.

Structural integrity is requi~ed to ensure that the containment will withstand the maximum pressure of 49.)~psig in the event of a LOCA.

The containment design pressure is 60 psig.

The measurement of containment tendon lift"offforce; the tensile tests of the tendon wires or strands; the examina-tion and testing of the sheathing filler grease; and the visual examination of tendon anchorage assembly

hardware, surrounding concrete and the exterior sur-faces of the containment are sufficient to demonstrate this capability.

The tendon wire or strand samples will also be subjected to tests.

All of the required testing and visual examinations should be performed in a time frame that permits a comparison of the results for the same operating history.

The Surveillance Requirements for demonstrating the containment's structural integrity are in compliance with the recommendations of Regulatory Guide l.35, "Inservice Surveillance, of Ungrouted Tendons in Prestressed Concrete Containment Structures,"

Revisio '984'l k&fR-.

7Q)

'he required Special Reports from any engineering evaluation of containment abnormalities shall include a description of the tendon condition, the condition of the concrete (especially at tendon anchorages),

the inspection proceduresi the tolerances on cracking, the results of the engineering evaluation, and the corrective actions taken.

PALO VERDE - UNIT 2 B 3/4 6-2 OCT 41 1%5

e

~,

l, I

Proposed Changes and Basis for Changes To Technical Specification Table 4.12-1 The proposed changes are on the attached marked-up pages.

The basis for the change is as follows:

This appears to be a typo, as Unit' Technical Specification has the word water listed.

),

~

(

I TABLE 4.12"1 DETECTION CAPABILITIES FOR ENVIRONMENTAL SAMPLE ANALYSIS LOWER LIMIT OF DETECTION (LLD)

WATER ANALYSIS (pCi/E)

Gross eta 4

AIRBORNE PARTICULATE OR GAS (pCi/ms)

0. 01 MILK (pCi/R)

~

FOOD PRODUCTS (pCi/kg,wet)

H"3 Mn-54 Fe-59 Co-58,-60.

Zn-65 Zr"95 Nb-95 I-131 Cs"134 Cs-137 Ba-140 La-140 I

2000" 30 15 30 30 18 60

0. 07

0. 05

0. 06 15 18 60 15 60 60 80 Note:

This list does not mean that only these nuclides are to be detected and reported.

Other peaks that are measureable and identifiable, together with the above nuclides, shall also be identified and reported.

Ifno drinking pathway exists, a value of 3000 pCi/R may be used.

P coaWc.r

"~If no drinking pathway exists, a value of 15 pCi/R may be used.

'eak-a;c

1 P

Arizona Nuclear Power Project

.P.o. BOX 52034

~

PHOENIX, ARI2ONA85072-2034 Director of Nuclear Reactor Regulation Mr. George N. Knighton, Chief Licensing Branch No.

3 Division of Licensing U.S. Nuclear Regulatory Commission Washington, D. C.

20555 September 30, 1985 ANPP-33610-EEVB/KLI

Subject:

Palo Verde Nuclear Generating Station (PVNGS)

Units 1, 2, and 3

Docket Nos.

STN 50-528(License No. NPF-41)/529/530 Containment Analyses Results Chapter 15 Reanalyses File:

85-056-026; G.l.01.10

Reference:

Letter to G.

M. Knighton, NRC, from E. E.

Van Brunt, Jr.,

ANPP, dated April 15, 1985 (ANPP-32401);

Subject:

Revised Chapter 15 Analyses

Dear Hr. Knighton:

The attached supplies the results of the following analyses:

FSAR Section 6.2 Event CBIT LOCA DEDLS 9.82 FT2 Hax ECCS 6.2 CZ'IT SLB 102K Power 8.78 FT2 Slot with Cooling Failure 6.2 6.2 CTi'IT"SLB 75/ Power 8.78 FT2 Slot with Cooling Failure CTMT SLB 50K Power 8.78 FT2 Slot with Cooling Failure 6.2 6.2 CTi'IT SLB 25K Power 8.78 FT2 Slot with Cooling Failure CPalT SL'B OZ Power 4.0 FT2 Slot with Cooling Failure ANPP committed to the reanalysis and.submittal of these events in the referenced letter.

I I

~

G.

M. Knighton, Chief

~

Containment Analyses Results Chapter 15 Reanalyses ANPP-33610 Page 2

The results are presented as proposed changes to FSAR Sections 1.9.2.4, 6.2.1 and 6.3.1.4.M and are all within NRC acceptance criteria.

Upon iVRC approval, these FSAR changes will be incorporated into the next PVNGS FSAR Amendment.

e The analyses were performed with the following parameter changes:

Containment Spray (CS)

Pump Flow Rate.

A lower CS pump flow rate of 3525 gpm, instead of 3740

gpm, was utilized to allow for additional pump operating margin over pump life.

Main Steam Line Break (iilSLB) Blowdown Data This data was revised due to the use of PVNGS specific as-built MSL pipe I.D. of '25.5 in. 'rather than 28 in. which is used in CESSAR.

Using the 25.5 in. diameter reduces the maximum break size from 8.78 2ft to 7.16 ft.

The reduced diameter also decreases the size of the OX power MSLB from 4.0 ft'o 3.0 ft'.

The 3.0 ft'reak is now the maximum break size which does not result in mixed flow out the break.

Please contact Mr. M. F. quinn, of my staff, if you should have any questions on this matter.

~ J EEVB/KLM/slh Attachment Ver> truly yeure, C:

C t

c.4C~~

E. E.

Van Brunt, Jr.

Executive Vice Presient

"'roject"'Director cc:

E. A. Licitra R.

P.. Zimmerman A. C. Gehr

~ e N

0 f

t I

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.

~

<<0 g

/

Edwin E.

Van Brunt, Jr.

, ~

~

'~ ~:,'i

/'if~ZZ (u

II7,'n Notary Public, My Commission Expires:

wig,Crtpwtqepj pvrji <~r:..g~r:j g co@7

0 I

f i

G.

W. Knighton Containment Analyses Results Chapter li Reanalyses ANPP-336yP Page 3

0 bcc:

D. B. Karner J.

G. Haynes R. M. Butler W. F. Quinn T. F.

Quan J. Vorees D. N. Stover P. F. Crawley LCTS Coordinator K.

W. Gross J.

H. Allen J.

R.

Bynum

0. J. Zeringue J.

D. Houchen S. Shapiro C. F. Ferguson N. Barnoski

PVNGS

FSAR STANDARD DES IGNS Containment Spray System (CESSAR Appendix 6A, Section 7.14/FSAR section 6.5.2.8 (RA) 7.1.4) 4 CSS actuation and flow delivery occurs on a pre-established schedule given in Section 6.3 of CESSAR Appendix 6A.

This schedule determines flow conditions in various analyses that take credit for a train of the CSS functioning to mitigate the MA>iN V hh consequences of postulated transients.

The~time to establish rated flow given in CESSAR Appendix 6A, Section 7.1.4, is 58 seconds after receipt of CSAS. ~ PVNGS,CSS rated flow will POP.

C".RCAY68 'TteAW 68 56<OWD> (T~<g

~, g. )-q),

be delivered in The exception to the CESSAR interface requirement for the CSS flow delivery time is only significant in the determination of the in-containment equipment qualification temperature parameters.

However, unde'r

~>P~b T~RC

~..

I ~ l4 the worst case >conditions (i. e. Nain Steam line~ft slot break at 102% power, loss of one containment spray train, and no loss of offsite power}, the calculated temperature transient of section 6.2.1.8 is bounded by the equipment qualification environment of table 3E-1 (sheet 1 of 7).

C-E has also committed to qualify safety-related equipment within their scope to the same environment.

By this analysis, the delay in delivery of rated flow to the CSS has been shown to have no adverse con-sequences under postulated transients.

1.9.2.4.2 Containment Spray System (CESSAR Appendix 6A, Section 7.13.14/FSAR section 6.5.2.8 (RA) 7.13.14)

The head loss requirements of CESSAR Appendix 6A are not applicable to'he PVNGS design due to the addition of the auxiliary spray headers.

The analyses of both train A and e

train B of the containment spray system have shown that, either train of the system will operate satisfactorily and meet the design flow requirements independent of the other train during all modes of operation.

The design flow includes the additional flow for auxiliary spray headers.

December 1980 Amendment 3

0

PVNGS-FSAR CONTAINMENT SYSTENS Table 6.2.1-1 POS "LAT D AC I EN S

FOR CONiTA NMEN1T DESICN (Sheet 1 OI 2)

C"nt inment Design Parameter Peak pressure/temperature Postul ted Accidents Considered Loss-of-coolant accidents (LOCA)

Double-ended ho" leg sic (DEHLS), 19. 24 ft2 area Double-ended suet'on leg slot (DESLS),

9. 82 f" are, Maximum ECCS flow 1

Double-en"ec sue~ion leg s'o (DESLS),

9.82 <<t

area, Min'mum ECCS flow Double-enced dischar"e leg slot (DEDLS),. 9.82 ft" a ea, Max'm;..

ECCS f'o~

Double-ence" dis"harge le, slot (DEDLS), 9.62 ft

area, Min'm ;

ECCS flow Main steam lire breaks (MS:B) 7./b 2ft slot area MSLB, 102"'

power with cooling train

='-ilure Qa/Q ft slot area MS B, 7SE powe with cooling tra'n failure 7-)4 ft slot area

NSLB, 5G5 power with cool'ng train failureft slot are

MSLB, 2S=.

. power with cooling train failure P~

2 ft slo area MSLB, 0.04 powe with cooling train failure

6. 2. 1-3 a

0

. t

PVNGS FSAR Table 6.2.1-2 CONTAINMENT SYSTEMS CALCULATED VALUES FOR CONTAINMENT PRESSURE PARAMETERS Parameter Design B~~js Calculated Value (psid)

Peak pressure Peak subcompartment pressure DEDL slot, Max.

ECCS Reactor cavity 350 in.2 DLG 99.8 External pressure loading Steam generator compartment 592 in.2 SLG Pressurizer compartment wall 161 in.2 surge line guillotine Inadvertent operation of the containment spray system 2,9. 4 72.4 2.6 a.

See table 6.2.1-1 for definition of abbreviations used.

February 1985

6. 2. 1-5 Amendment 14

0

PVNGS FSAR CONTAINMENT SYSTEMS the design pressure (60 psig), and the caaaulated peak pressure Ae. 5 of the as-constructed design>(+&~ psig) results in a design 1

margin of. approximately 20%.

6.2.1.1.1.2 Containment Internal Structures Accident Conditions.

The RCS breaks defined in CESSAR Section 3.6.2 are analyzed and form the design basis for the loads on containment internal structures and equipment.

The simultaneous occurrences are the same as discussed in section 6.2.1.1.1.1.

The design pressures for the reactor cavity, steam generator

'I and pressurizer compartments are given in table 6.2.1-3.

The cu rently calculated values for the as-built design are listed in table 6.2.1-2.

6.2.1.1.2 Design Features The design features o'f the containment structure and internal structures are provided in sections 3.8.1 and 3.8.3, respec-tively.

6.2.1.1.2.1 Protection From The D namic Effects of Postulated Accidents.

The containment structure, subcompartments, and engineered safety feature systems are protected from loss of safety function due to the dynamic effects of postulated accidents.

Containment design has provided separation and inclusion of barriers, restraints and supports when required to protect essential structures,

systems, and components from accident generated missile, pipe whip, and jet impingement forces.

The detailed criteria, locations, and descriptions of devices used for protection are given in section 3.6.

September 1980 6.2.1-7 a

I

.;,',,:.".';:..Amendment 2...

~

)

PYNGS FSAR CONTAINMENT SYSTEMS Table 6.2.1-4 MASS AND ENERGY RELEASE FOR CONTAINMENT PEA< PHESSU.":E/TEMPERATURE ANALYSIS (Sheet 1 of 4)

A.

Most severe hot leg break Brea'k type:

Double-ended hot leg slot break Pipe ID, 42 in.

Break area 19.24 ft Accident Phase Time (s)

Mass Release Rate

( ibm/s)

Energy Flow Rate Million Btu/s Enthalpy (Btu/1 bm)

Reactor Y.esse 1 Pressure psia Blowdown Refe-to CESSAR Table 6.2.1-10 a

~

70 psia containment backpressure case b.. Recirculation begins c

~

d.

e.

Safety injection realigned; 50'4 to hot legs and 50% to cold legs.

Excess hot leg safety injection water carries significant decay and sensible heat to containment sump.

RCS sensible heat addition to containment completed.

Includes direct floor spillage for discharge leg break.

f.

Fe~r ~ to table 6.2.1-~ for PVN specific mass/

energy lease data used for equipment qua ication peak temperature n

sis in section 6.2.1.8.

February 19=.=

o. 2. 1-13

'I Amendment 14

0, I

r/

PVNGS FSAR CONTAINMENT SYSTEMS Table 6.2.1-4 MASS AND ENERGY RELEASE FOR CONTAINMENT PEAK PRESSURE/TEMPERATURE ANALYSIS (Sheet 4 of 4)

PEpf ~cf Ni7H INSF.5 Pipe ID: ~ in. Break area: ft Reactor power level: 1024 Time (s) Mass ase Rate (ibm/s ) Enthalpy tu/ibm) Energy Release Rate (million Btu/s) - Refer to CB'.MAR Table gp/ l~('g cJt7H I~ Amendment 14 6.2.1-16 .;; February 1985 l ~ ~, TlIGe (s) Mass Release Rate (1 bm/s) Enthalpy (Btu/ibm) Energy Release Rate (million Btu/s) 0.0 0 ' O.c 0.6 2.0 5.0' n 7.C 8'.0 1' 12.0 .13 ~ C 14.C 15.0 2A 0 25.0 3C.O, 35.0 40.0 45.0 5G.O 60.0 70.0 80.0 90.0 100.0 110.0 120.0 130." 14'i. 0 150.0 160.0 l7o.o 150.0. 190.0 %0.0 8571.5 7155.9 7059.9 7043.0 6'55=.1 6c1c 3 6715.7 6c C p 61co 0 4C54 C75' 46:5.6 4 5C.C 4C C 25:3.2 2c=c c 2537.0 25 7.6 2475.2 2452.5 23pc.s 2173.7 2076.4 1995.9 1928.1 186c 1 1813. 7

l. 15.1 1625.3 1535.9 1C60.6 13ro 9

1310.9 123C.3 1156.2 1072.0 971,4 831.6 624.9 354.0 25.0 3.0 lie.O 1185.9 1186.4 118=.8 118..2 llc7.5 1188.9 1150.0 115i.l 1152.1 1193.2 1153.8 1154.7 lloc 1 lice' 1156.0 1196.4 1155.7 1198.3 1159.3 -'1200.6 1201. C 1201.7 12"2.3 1202.5 1203.4 1203.5 1204.0 120C.3 1204.6 1204 ' 1204.7 1204.8 1204.3 1203.4 1202.1 1199.4 1241.5 1262.4 1261,3 1G.640 8.5"3 8.423 1C 1C 8.2 7

7. 98.5 7

7 ~ 5.7=. C C ~ J 5.572 5 crc 5 3.11". 3 A 3.G31 2 oc. O1C ~ ~

2. 764 2.6C7 2.493 2.355

2. 317 2.245 2.181 oc 1

85C 1.7c5 1.55 1.579 l.u7 1.353 1 Zcl 1.169 0.9957 0.7489 0.4395 0 031C6 0.003754 Ti7Qe (s) Mass Release Rate (ibm/s) Entha1 py (Btu/ibm) Energy Release Rate (million Btu/s) 0.0 0.2 0.4 0.6 0.8 1.0 2.0 3.0 4.0 5.0'.0 7.0 8.0 9.0 10.0 11.0 12.0 13.0 14.0 15.0 20.0 25.0 30.0 35.0 40.0 45.0 50.0 60.0 70.0 80.0 90.0 10C.O 110.0 'l20.0 130 c0 140. 0 150.0 '160.0 170.0 180.0 190 A 200.0 98SS.O 7185.5 7031.2 6901.0 6790.8 669c.4 6353.4 6118.6 5938.2 4532.2 4595,1 l63c.6 466. 8 c92.9 '608.8 26Z?.8 2633.5 2637.8 2633.2 2636.7 2524.9 2l08.1 2321.0 2225.7 2146.5 2078.3 2009.6 1862.3 1764,4 1649.8 1509.2 1341.4 1182.3 llCS.7 ~ 10CC.S 867.3 707.1 53\\.2 383.9 193.9 6.6 3.$ 1189.9 1190.8 1191.5 1192.1, 1192.6 1193.1 1154.6 1l95.7 1196.5 119c.9 1195 ' 1195.3 1195.3 1195.2 1194.6 1194.5 1194.6 1194.6 1194.l 1194. 3 1195.7 1197.2 1198.2 1199.2 1199.6 12CC.5 1201.2 1202.3 1202.7 . 1203.8 1203.9 1204.7 1204.4 " 1204.7 1203.4 1202.6 1203.1 1196.2 1190.2 1242.9 1289.4 1292.1 11.73 8.561 8.378 8.227 8.099 7.988 7.586 7.316 7.10S 5.420 5.491 S.sll 5.578 5.609 3.117 3.133 3.146 3.151 3.145 3.149

3. 019 2 ~ 863 2.781 2.669 2.575 2.495 2.414 2.263 2.122 1.986 1.817 1.616 1.424 1.332

'1.204 \\.043 0,8486 0.6354 0,4569 0.24 10

0. 00851

0. 00491

PVNGS 2SAR CONTAINMENT SYSTEMS Table 6.2.1-7 ENGINEERED SAFETY FEATURE SYSTEMS OPERATING ASSUMPTIONS FOR CONTAINMENT PEAK PRESSURE ANALYSIS (Sheet 1 of 2) System/Item Full Capacity Value Used for Peak Pressure Analyses Passive Safety Injection System Number of safety injection tanks Pressure

setpoint, psig 3

Volume, ft /tank Active Safety Injection Systems High-.pres~use saf ety injection a Number of lines Number of pumps NLk Flowr a te. gal/min/pump Low-pressure safety u e injection Number of lines Number of pumps HN. Flowrate. gal/min/pump Containment Spray System Number of lines Number of pumps Number of headers (c)

Flowrare, gal/min/pump f(~ VO

'~ rated. flow, seconds AF'Tag Ev (Offsite Power Available/ ~ Loss of Offsite Power) Loss of coolant accident Main steam line break accident 600 1927 1130 5000 HT tNiTiAT! 600 1927 2 1/2( ) 1130 2 1/2'" 5000 1 1 8I(9I s%< February '1985 6.2. 1-19 Amendment 14 PVNGS FSAR. CONTAINMENT SYSTEMS Table 6.2.1-7 ENGINEERED SAFETY FEATURE SYSTEMS OPERATING ASSUMPTIONS FOR CONTAINMENT PEAK PRESSURE ANALYSIS (Sheet 2 of 2) System/Item Full Capacity Value Used for Peak Pressure Analyses Heat Exchangers Shutdown heat exchangers Type Number 2 Heat transfer area, ft Overall heat transfer coef ficient, Btu/h-ft F Flowra tes: Recirculation side, gal/min Exterior side, gal/min Source of cooling water Shell and U-Tube 10,840 372 9d Ci$" 14,400 Essential cooling water Shell and U-Tube 10,840 285 14,400 Essential cooling water a. From CESSAR Section 6.3 b. 1 = minimum ECCS; 2 = maximum ECCS c. Net flow during injection phase; flow increases to gal/min following closure of miniflow bypass valve on recirculation actuation signal (RAS). both maximum and minimum ECCS performances were evalu-ated. For the containment heat removal

systems, mini-mum system capacity is conservative for calculating containment peak piessures.

passive heat sink data is provided in table 6.2.1-8. Part A of the heat sink table is a detailed list of the geometry of each heat sink and part B describes Amendment 14 6.2.1-20 .4 February 198S PVNGS FSAR CONTAIRMNT SYSTEMS components would be cooler than the design maximum temperature; a few regions, notably the reactor cavity walls, may be warmer. The warmer heat sinks comprise a small area

and, as a result, the average heat sink temperatures would initially be less than the design maximum.

Minimum heat sink areas (i.e., nominal minus tolerance given in table 6.2.1-8) are used for the short.-term and long-term containment pressure-temperature trans ent analysis of primary or secondary pipe ruptures inside containment. Most individual-structural or component heat sinks have some uncertainty associated with the exposed surface area. The minimum area was used to provide the most conservative.assumptions. Table 6.2.1-8 part C, lists the thermophysical properties used in analyses. Metal, concrete, and protective coating properties are typical values for the temperature range observed. Blowdown mass and energy release rates for LOCA ~ /gal d' 6. I dd'6dd6dd computer code is presented for the most severe break area at each break location in table 6.2.1-4. Assumptions made regarding closure times of secondary system isolation valves and single active failures for MSLB analyses are discussed in section 6.2.1.4. 6 C.'ethods of Analysis r The containment pressure analyses are performed using the Bechtel COPATTA computer program that was derived from the CONTEMPT program written for the AEC Loss-of-Fluid Test (LOFT) program. HSLB rddCrS~ fagg ad P ~ictuSSrd l< SfC7i p< $.g./.Q. 0 PANGS FSAR CONTAINNENT SYSTEMS transferred between the liquid and vapor regions by boiling, condensation, or liquid dropout. Evaporation is not considered. A convective heat transfer coefficient can be specified between the sump liquid and atmosphere vapor regions.

However, since any heat transfer in this mode is small, a conservative coefficient of zero is generally assumed.

Each region zs assumed homogeneous, but a temperature difference can exist between regions. Any moisture condensed in the vapor region during a time increment is assumed to fall immediately into the liquid region. Non-condensible gases are included in the vapor region. Accident Identification and Results The containment pressure and temperature response and sump water temperature response versus time are given in figures 6.2.1-1 through 6.2.1-6 for the most severe LOCA breaks and the most severe NSL breaks. It, has been demonstrated in reference 1 where main steam line .breaks produced a high degree of superheat that typi-cal safety-related equipment surface temperatures remained at or near the containment saturation tempera-ture as a result of the short time frame at superheat conditions.due to spray actuation. Pipe break locations, break areas, peak pressures and temperatures, times of peak pressure, and total energy released to containment"are summarized in table 6.2.1-9 fo'r each LOCA and IISLB analyzed. Based on the results presented in these tables, the double ended discharge leg slot break LOCA with maximum ECCS was identified ag t e pipe break with the highest peak pressure ( ". psig) which is below the design pressure value of 60 psig. 6.2.1"31 Table 6.2.1-9

SUMMARY

OF CALCULATED CONTAINMENT PRFSSURE AND TEMPFRATURFS Loss oi Coolant hccldent Results Peak pressure, psig Peak temperature,

. F 0

Time of peak pres-

sure, seconds Energy released to containment up to the end of blow-

down, 106 Btu Hain Stream Line Brcak Results Peak pressure, psig Penk temperature, F

0 Time of peak pres-

sure, seconds Energy released to containment vp to the end of blow-

down, 106 Btu DEIILS 19.24 ft 45.1 273 8.6 376.00.'.lb 1021 Power 8,78 ft2 slot with Cooling Failure

>~Re.5

~ lga Mo.O DESLS2 9.82 Hnx ECCS

46. 3 298 87 378.10 51 Power

.78 ft2 slot VHh Cooling Failure

.I Ml 3 DESLS2 9.82 Hin ECCS 46.5 300 87 378.10 01 Power 8.78 ft2 slot Cooling Fni lure DEDI.S2 F 82

~ Hnx ECCS

~%5

~23 I

382.90 Power 8.78 ft2 slot Cooling Failure DEDLS2 9,82 Hin ECCS 45.1 297 500 382i90 Ol Power 4.QO ft2 slot w th Cooling Fai)ure

CONTAINMENT SYSTEMS Figur'es 6.2.1-7 through 6.2.1-9 are plots of the con-tainment condensing heat transfer coefficient versus time for the most severe RCS discharge and suction leg breaks and secondary coolant system breaks.

6.2.1.1.3.2 Lon -Term Containment Performance.

Long-term analyses of the worst case pump discharge leg break, and the worst case pump suction leg break were performed to verify the ability of the containment heat removal system (CHRS) to main-tain the containment below the design conditions.

These evalu-ations were based upon conservatively assumed performance of the engineered safety features.

The CHRS long-term operating mode is assumed to include one containment spray train.

The containment pressure-time responses for the DBA for the pump discharge leg and the pump suction leg cases out to 6

10 seconds (11.6 days) are shown in figures 6.2.1-3 and 6.2.1-4. for the ESF performance mode outlined in table 6.2.1-7.

The containment pressure-time response for the highest pressure MSLB case (0% power) is shown in fig-3 ure 6.2.1-6..

-The-8MB. analysis shows that by 10 seconds the containment pressure is reduced to~psig which indicates that the containment pressure will be reduced below 50% of the peak pressure within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

No further analysis is performed for the NSLB since after isolation and blowdown there is no further energy input to containment.

The maximum pressure of

//8 4i..& psig occurs at H% seconds in the 0% power MSLB case.

The IQC.y energy distributions. in the containment versus time are shown in figures 6.2.1-10 through 6.2.1-12 for DBA LOCAs and NSLBs.

Mechanisms of energy removal from and transfer within the containment are 'addressed.

Included are the vapor energy (steam plus air),

sump (liquid) energy, energy contained in

,heat sinks, energy removed by the shutdown cooling heat exchangers, and energy transferred by sprays from the vapor to the sump.

6.2.1-33

~

i 0

PVHGS FSAR CONTAINMENT SYSTEMS Table 6.2.1-10 ACCIDENT CHRONOLOGIES (Sheet 3 of 7)

C.

Horst case discharge leg break Break type:

Double-ended discharge leg slot break Maximum ECCS

. Time (s)

Event 0.0 Break occurs

14. 4 15 19.3 19.3 217 Start core flood tank injection Peak containment pressure (blowdown), 43.0 psig Start ECCS injection phase End blowdown Start spray injection End of core reflood 500 2I17 Peak containment pressure subsequent to end of blowdown, ~ psig 9'f. 5 End of steam generator energy release:

post reflood End of ECCS injection; start ECCS recirculat'ion

'57bo 50&9 5400 86400 10 End of spray injection; start spray recirculation Containment pressure at 30 psig (one-half 60 psig design value)

ECCS realigned:

504 to hot leg and 50'o cold leg Addition of RCS sensible heat completed Depressurization of containment, 2.1 psig Yebruary 1985 6.2.1-37 I

Amendment 14

II ~

~

~ ~

0

~

~.

~

~I

~

PVNGS FSAR CONTAINMENT SYSTEMS Table 6.2.1-10 ACCIDENT CHRONOLOGIES (Sheet 5 of I)

DE Horst case main steam line break (pressure)

Break type:

OR power MSL slot break Loss of cooling train (cont)

Time (s)

Event Main steam isolation valves closed~

Main feedwater isolation valves closed (see section 1.9.2.4.10)

Containment spray pump at full speed Containment spray at full flow initiated inside containment building 37-.2 ~-

. Peak containment te poerature of ~occurs Peak containment pressure of ~~psig occurs

+/. 4 Blowdown ends February 1985 6.2.1-38A

...'.Amendment 14

0'

I

~

0

~I

~

0 0 0

~

0 0

~ ~

~

0 ~

0

~

0 r 0

o

~ ~

0 II 0

0

~

0

~

~ ~

~ 0

~

0

~

0

~ ~

~

~I

~

~ ~

~

0

~

~ ~

0

L VN4b l"bAR CONTAINMENT SYSTEMS Table 6.2.1-10 ACCIDENT CHRONOLOGIES (Sheet 7 of 7)

E.

Worst case main steam line break (temperature)

Break type:

102t power MSL slot break Loss of cooling train (cont)

Time (s)

Event

~B,o 8,45 l3. lo5 CSAS generated Hain steam isolation valves closed Main feedwater isolation valves closed (b)

(see section 1.9.2.4.10) ee 8a 44-84

~ l12..

~ g.OQ Containment spray pump at full speed Containment-spray at full flow initiated inside containment building SS8.5 F

.Peak containment temperature of ~~ occurs H I. 0

'eak containment pressure of~ psig occurs Blowdown ends February 1985 6.2.1-39 Amendment 14

0

Table 6.2.1-11 REACTOR CONTAINMENT BUILDING ENERGX DISTRIBUTION (Sheet 6 of 11)

Worst case discharge leg break Break type:

Double ended discharge leg slot break,. maximum ECCS Break area:

9.82 ft2 Energy

( 10 Btu) 6 Energy Description Reactor coolant system water internal energy Safety injection tank water internal energy Energy stored in core Energy stored in RV internals Energy stored in RV walls Energy generated during shutdown from decay heat Prior to LOCA 357.992 41 ~ 846 30.299 37.386

88. 602 At Peak Prcssure Prior to End of D lowdown 18.888 29.207 13.923 34.959

88. 531 5.188 End of:

B lowdown 14.243 24.869 13.741 33.948 88.400 5.954 At Peak Pressure Aftrr End of D lowdown

>~~5 l 1

MS

'Ll. 3r>

0~98.

9I

'1 3.S5Q End of Core Rcflood 40.385 0.0 6.760 28.811

86. 937 36.621 End of Post Rcflood 31.239 0.0

5. 142 23.550 84.355 70.364 24 Hours After LOCA (86400

~ )

26.09 0.0 0.0 0.0 0.0 2955.30 Energy stored in pres-

surizer, primary piping,

valves, and pumps Energy stored in steam generator tubes 124.294

32. 180 123.283

29. 669 122.810

29. 815 40+rfEHr 122, 816 10'l.?, 3 I 17.849 100.178

15. 445 0.0 0.0

t

~

~ ~

~

r

~

~ ~

~

0

~

~,

~

~ ~

Tab'le 6. 2.1-11 RHACTOR CONTAINMH BUIL'DTJJG ENERGY DT, TR'fBlJTION (Sheet 8 of 11)

~(a)

D.

llorst case sec ndary system l>rr <<k (co>>tainment temperature)

Bx eal: type:

102. 07, power ma in steam l i.n>> hreak Break ~rca 8

70 ft~

Energy

.(10 Btu) 6 t>>

~

~ ~ ~

Energy stored in piping, valves, Envrily Ul:ored in pressurizer, primary and pumps steam generator tul>es Energy Description Rea~.l or coolant system water internal energy Safety injection tank water internal energy Energy stored in core Energy stored in RV internals Energy stored in RV metal Energy generated during shutdown from decay heat Prior to MST.B 357. 992 NA 30.299 NA

0. 000 250. 202 32..100 At Peak Pressure Prior to End of: 13lowdown m~e+

2$ 4 pro NA tl.

F53'A NA g~-F70 A l.6~

P Jz.l ZE pl -7g3 End of Blowdown

~~8-4 2V~- Fg(

NA

]J

~r6 NA NA c0 5t<

~H8-

~

a

~

~ ~

~

~ I

~

, ~

~

e e

~

~ ~

~

9.0 Table 6.2.1-11 REACTOR CONTAINMEN 'BILDING ENERGy DTSTRIB()TION

'Sheet 11 of 11)

(a) l'..

Worst case s< condary system break (containment prrssure)

Break type:

0%

~ower main steam line break Break area:

.0 f:t2 Energy (10 Btu) 6 Energy Description Energy stored in steam generator secondary walls Secondary coolant internal energy (in steam generators)

(b)

Energy content of RCB atmosphere P

~ Energy content of RCB internal structures(c)

Energy content of recirculation intake water (sump)

Prior to MSLB 157.441 369. 750 6.73 0.0 0.0 At Peak Pressure Prior to Fnd o f B lowdown 1'3

&%~<1 (4o 2'fS Fnd of Blowdown 27-7(

Energy <.<<ntent of RWST water Feedwat';.

l.o steam generator l Feedw'.I

(.o steam generator.

2 349.55 0.0 0.0 0.0 o Ao 3

0.0 o 477 rl

PUNG' SAR CONTAINMENT SYST'cc.

Table 6.2.1-25 (a)

NISCELLANEOUS AND ADDITIONAL ENERGY RELEASES Time (sec)

Energy (Btu/h) 0 500 0

500 86,400 86.400 20.8956 E6 20.8956 E6 0

a.

All misce'aneous ene gv added af te blowdcwn ends (500 sec),

up to 1 day (86. 400 sec)

Secondary System Pipe Ruptures Inside Containmen-Refer to ESSAR Section 6.2.1.4.

The MSLBs analyzed for con-tainment pre re/temperature were the slot breaks listed in CESSAR Tables 6.2.

11 through 6.2.1-20.

The guillotine ruptures for the same s'

break and power levels were evaluated as being less severe than th omparable slot breaks.

In addition.

PVNGS-specific mass/ene release data for the 1024 power MSL slot break with a 9.6 cond feedwater isola-tion valve closure time used for the in-co ainment equipment qualification peak temperature analysis are pr ded in table 6.2.1-25A.

Qppl'p~ ~/g

/noir<

.c Februa r ~ ii=5 6.2-1-96I

,, Amendment 14

I

INSERT "C" The MSLB methodology used for the PVNGS'analysis is the same as CESSAR Section 6.2.1.4 except as follows.

'The PVNGS containment free volume of 2.6 million cubic feet was used in the calculation of the time to reactor trip and MSIS on containment high pressure rather than the 3.7 million cubic feet used in CESSAR.

The PYNGS specific steam line arrangement was used for determining flow from the intact steam generator to the containment prior to isolation.

The steam line model considered choking at the MSIY throat and cross connect piping as well as at the venturi throat.

The MSIV and MFIV closure times were 4.6 and 9.6 seconds respectiveIy.

The Palo Yerd specific mass and energy release data for'he-most severe HSLB containment pressure and temperature cases are given in table 6.2.1-4.

~

ICTUS PVNGS FSAR Table 6.2 1-25A CONTAINMENT SYSTEMS ENERGY RELEASE FOR CONTAINMFNT EQUIPMENT

'UALIFICATIOHPEAK TEMPERATURE ANALYSIS (Sheet 1 of 2

Break type:

Pipe ID:

Break area:

Reactor power level:

Isolation valve closure tines:

Main steam line slo 28 in.

8.78 ft2 102%

4.6 sec MSIV 9 ~ 6 sec MFIV break Ti me (s) 0.0 0.2 0 '

1.0 2 '

3.0 4

~ 0 5.0 6.0 7.0 8.0 9

~ 0 9.5 10.0 10.5 11.0 11.5 12 13

~ 0 15.0 20.0 Mass Release Rate (ibm/s) 9833.75 7303.17 7039

~ 74 6849.83 6524.69 6301. 20 6127.74 4660.96 4759 '6 4843.

4 492

.50 4

8.01 016. 57 2853.53 2864. 79 2870.74 2871.49 2867.25-2858 '5 2844.85 2767 '5 2559.88 Enthalpy (Btu/ibm) //

1190 F

04'190.89 119' 87 1 42.71 194.18 1195 '1 1195.99 1195.40 1195

~ 02 1194.57 1194.13 1193

~ 72 1193.54 1192.92 1192.77 1192.67 1192.65 1192

~ 68 1192. 77 1192 '1 1193.77 1196.08

/y Energy Release s

Rate (million Btu/s) 11.702518 8.695786 8.390426 8.169867 7.791653 7 '31249 7 '28734 5.571987 5.687764 5..786444 5 875717 5.954272 5.987483 3.404022 3 '17024

, 3.423858

~ 424678 3.'419704 3 '09/35 3.393662, 3.303575 3.061811

PVNGS FSAR

~36 tE CONTAINMENT SYSTEMS Table 6.2.1-25A MASS AND ENERGY RELEASE FOR CONTAINMENT EQUIPMENT QUALIFICATION PEAK TEMPERATURE ANALYSIS (Sheet 2 of

)

ime

(

Mass Release Ra te (ibm/s )

Enthalpy (Btu/ibm)

Energy elease te (mil on Btu/s) 25.0 30.0 35.0 40

~ 0 45.0 50.0 60.0 70

~ 0 80.0 90.0 100.0 110.0 120

~ 0 130.0 140.0 150.0 155.0 160.0 165.0 170.0 180.0 185.0 190/

1 4.5 2445.21 2374.74 2284 '5 2206.28 2142 '9 2075.62 1948.28 1828.22 1720.88 1610.06 1492.90 1361 '7 s

1249. 5lr 1177.

6 1044.42 970.36

/

~~903

~ 62 830.21 751.09 676.99 512.9 380 '

261.4 0

1197.29 1198.00 1198.90 1199 '4 1200.2 1200 80 12 IL.84 02.70 g 1203 '4 1203.87 1204.24 1204.41 1204.36 1204.22 1203 F 85 1203.09 1202.46 1201.57 1200.37 1198.94 1200.4 1227.3

'239.5 1240.3 2.927615 2.844945 2.738457 2.646736 2.571811 2 '92405 2.341530 2.198796 2.070804 1.938305 1.797816 1.639290 1.504854 1.417440 1

~ 305478 1-167428 1.086565 0-997553 0.901587 0.811673

',0.615700 0.467100 0 ~ 324000 0

tegral:

329100 ibm 395.014 million Btu

PVNCiS F SAR CONTAINMENT SYSTEMS 6.2.1.6 Testin and Ins ection Testing and inspection'requirements for the containment are discussed in section 6.2.6.

No other testing of the contain-ment structure is planned or required.

Testing and inspection requirements for other engineered safety features that interface with the containment structure are discussed along I'with the-applicable system descriptions.

6.2.1.7 Instrumentation A

lications The containment pressure is measured by independent pressure transmitters located at widely separated points outside the containment.

Refer to section 7.3 for a discussion of pres-sure as an input to the engineered safety features actuation system (ESVAS).

Refer to section 7.5 for a discussion of the display instrumentation associated with pressure.

The containment airborne radioactivity is monitored by the airborne radioactivity monitoring system, discussed in section 11.5.

Hydrogen concentration is monitored in the containment by the hydrogen monitoring system.

discussed in section 6.2.5.

Temperature sensors are positioned at appropri ate locations throughout the containment.

The temperature js displayed in the main control room along with high-temperature alarms.

6.2.1.8 uglification Parameters for In-Containment Safet Related E ui ment In-containment safety-related equipment required to operate Post-MSf.B. is qualified to the Main Steam line break design basis accident environment as specified in table 3E-,l (sheet 1 of 7).

This environment bounds the calculated pressure-time and temperature-time response of figure 6.2.1-20.

The transient conditions shown in figure 6..2.1-20 are the result of the con-tainment analysis for a

1024 power.

.78 t

, slot-type Main Steam 2

line break, with the loss'f one con ainment spray train and no 7 l4 December 1980 6.2.1-97 Amendment 3

PVNGS F SAR CONTAINMENT SYSTEMS loss o

offsite powers It differs from the analysis of the Ma:.. Steam Line break of section 6.2.1.1.3.1 in that

/

~

~ r

~ ~

V

~ ~

AA

~ I

~

w 1

~ ~

C 4

~ ~ e the equipment qualification analysis utilizes 8% condensate re-evaporation as allowed, by Appendix B of NUREG-0588 A chronology of events for (5) this main steam line break analysis is provided in table 6 '.1-27 'ombined e ffects o f the PIGS-speci fi e irg of the 1025 MSi "t break compared with reviou equipment qualification con~

'.i,enr. temperature analysis based on the CESSAR mas ene-gy se data for the same break used in d'or. 6.2.1.1-3.1 is a 1.3=

uc. ion.in p'eak contai nt vapor temperature (368.5 vs 369-8E)'.an 4.5 psig crease in peak containment pressure (45.6 vs 41.1 psig In-containmert safety-related equiprent required tc ope ate post-LOCA is qualified to the LOCA design basis accident environ-ment as specif ied in table 3E-1 (sheet 1 of 7).

This environ-ment bounds the calculated pressure-time anc temperature-time response of figure 6.2.1-3 which shows the worst case LOCA transient as discussed in section 6.2.1.'"1.3.1.D.

) ~ I

~'

k'VNGS k SAR Table 6.2.1-27 CONTAINMENT SYSTEMS ACCIDENT CHRONOLOGY FOR CONTAINMENT EQUIPMENT QUALIFICATION PEAK TEMPERATURE ANALYSIS (Sheet 1 of 2)

Break type:

102K power MSL slot break Loss of one cooling train Time (s)

Event 0.0 Break occurs Containment pressure reaches safety injection actua-tion signal (SIAS) analysis setpoint (5 psig)

Containment pressure reaches reactor trip and mai'n steam iso tion signal (MSIS) analysis setpoint

(~sig)

SIAS generated Reactor trip sig'nal and MSIS generated~~~

Turbine admission valves close Reactor trip breakers open~~~

7I3 Containment pressure reaches containment spray actua-tion signal (CSAS) analysis setpoint (10 psig)

CSAS generated E.s5 s3 C

Main steam isolation valves close~i Main feedwater isolation valves closed~i>)

a.

Value used in pipe break blowdown analysis.

b.

Events based on time'(3 '

seconds) to reach reactor trip and MSIS for setpoint of 6 psig used in pipe break blowdown analysis.

February 1985 6.2.l-gg Amendment 14

PVNGS FSAR CONTAINMENT SYSTEMS Table 6.2.1-27 ACCIDENT CHRONOLOGY FOR CONTAINMENT EQUIPMENT QUALIFICATION PEAK TEMPERATURE ANALYSIS (Sheet 2 of 2)

Time (s)

Event iceS pump loaded on essential bus CS pump at full speed CS headers filled and full containment spray flow established gz z.9C Peak containment vapor temperature of RW3.5'F occurs Peak containment pressure of~~ psig occurs Failed steam generator dryout occurs, blowdown ends 6.2.1.9 References 1.

Shoenhoff, H. M. and Braddy, R. N., "Containment and Safety-Related Equipment Transient Temperature Analysis 2 ~

Following a Main Steam. Line Break," Bechtel Power Corporation, May 1975.

BN-TOP-4, Rev. 1, Oct. 1977, Subcompartment Pressure and Temperature Transient Analysis, Bechtel Power Corporation, San'Francisco, CA.

Amendment 14 6 '-3--100 b'av' February 1985

0

VAPOR TMAX PMAX VAPOR VAPO EMPERATURE 250 LI L'0 Iz z

C 20 z0V 10 C

4u SUMP WATER 1 SPRAY TRAIN ON C~ Sl SEC.

TMA'x 23' Svh TEh

'P WATE R IPE RATURE R E CIRC ULATI Oh FROM BLDG.

SUMP BcGINS 4 2060 sEc (2 H.P. Sl PUI.IPSI SUhrP 1'VATER VAPOR END RCS SENSIBLE HE AT ADDITION 8 86400 Scc.

124 HR) 200 c

D 150 C

C4 I

0 1

10 10 10 10 10 oe TIME FOLLOWINS BREAK (SECONDS)

I I

I I I I I I

I I I I I

/

I I

I

> I I I I

I I

1 I I I I I

I I

I I I I I

I I >t I

0

. REFL~ ~t~ 6<'T~~

p<oVo~W FiCo~

Palo Verde Nuclear Generating Station FSAR CONTAZNMENT PRESSURE AND TEMPERATURE RESPONSE g

DOUBLE-ENDED DZSCHARGE LEG SLOT BREAK, MAXZMUM ECCS Figure 6.2.1-3

72 PM AX 63.U"PSIA CI 271.5 SEC 360 56 CA lU CK I

cf a

32 CI SUIh'p TMAX 296.2 F 691 SEC I

4.I Oq TMAX 230'5 F

@500 SEC 1 SPRAY TRAIN ON t

691 SEC RECIRCULATION I2'117 SEC 320 280 240 I

200 I60 16 120 10 102 I

I I

I II 10 I

I II I

10 105 I

I I Bp

,06 TIME FOLLOWING BREAK (SECONDS)

PROPOSED Pah Verde Nuckar Generating Station FSAR CONTAINMENT PRESSURE AND TE."tPERATURE RESPONSE, DOUBLE-ENDED DISCHARGE LEG SLOT BREAK, MAXZMUIt ECCS Figurc 6.2.1-3

0

P 6 102M POWER VAPOR TEMPERATURE VAPOR T

~ 401 F

0 80 SEC 400 50 40 Q

Pn 30 Iz 20 I20U 10 SUh! P TEMPERATURE ESSUR=

SUh!P T

~ 255 5 F X

0268 C

MAX 8 154 SEC 1 SPRAY TRAIN ON 8 80 ScCONDS 4

C 30M C

I 200 100 0

10 I tel I

I I

I 100 TIME FOLlOINING BREAK ISECONDS)

'I000 I

I I

10000

.. p:C rgCi Palo Verde Nuclear Generating Station FSAR CONTAINMENT PRESSURE AND TEMPERATURE

RESPONSE

MSLB (SLOT)

WITH LOSS OF ONE CONTAINMENT COOLING TRAIN (Sheet; 1 of 5)

Figure 6.2.1-6

f

440 P 102M POSER PMAX

~ 55.

PSIA

0) 172 SEC 57 52 TMAX~ 398.5' 84 SEC 47 320 CL 280 D

I 240 MAX 6 320 SEC 0

C 37 I

32 g 200 1 SPRAV TRAIN ON e'84 SEC 22 120 IO 100 TIME FOLLOWING BREAK (SECQNOSI 1000 4

~

" Figure 6.2. 1-6

I

~

i

440 875'X POWER PMAX 54.1 PSIA

@189.5 SEC 55 400 360 TMAX CI 84 SEC 45 320

" 280 I

240 MAX 6 330 SEC cf 40 o.

35 c

I cf 3D ~

47 20D 1 SPRAY TRAIN ON C'4 SEC 25 20 120 1

~

I I

~

10 100 TIME FOLLOWING BREAK (SECONDSI 1000 I

I

~

15 10000 PROPOSED Palo Verde Nuclear ( eneratinp Stat>on FSAR CONTAINMENT PRESSURE AND TEMPERATURE

RESPONSE
MSLB (SLOT) WITH LOSS OF ONE CONTAINMENT COOLING TRAIN (SHEET 2 OF 5)
Figure 6.2.3.-6
'E
0 60% POWER VAPOR TEMPERATURE VAPOR TMAX 398.0 F 4 80 SEC 400 U
30 9
2 20 z
CI?0 10 SUMP TEMPERATURE PRESS SUMP TMAX 5'7 F 4 312 SEC MAX 4 197 SEC 1 SPRAY TRAIN ON 4 80 SECONDS 300 L
ICC 200 I
TIME FOLLOWING BREAK ISECONDS) 10000 pRDP0 s~
5 lGuRE Palo Verde Nuckar Generating Station FSAR CONTAINMENT PRESSURE AND TEMPERATURE.
RESPONSE
MSLB (SLOT)
WITH LOSS OF ONE CONTAIMKNT COOLING TRAIN tSheet 3 of 5) figure 6.2'-6
0
440 P SW POWER PMAX 53.2fPSIA
@209 SEC 55 400 TMAX 394.6~F CI 84 SEC 45 320 CL
~
2SD 240 200 e
TMAX 252.9F 5I 350 SEC 1 SPRAY TRAIN ON 0 84 SEC 40 cL CC 35 o-Ia
~e 30 C)
C'5 160 20 120 1
10 100 TIME FOLLOWING BREAK (SECONDS) 1000
~
~
15 10000 PROPOSED Palo Verde Nuclear Generatinp Station FSAR CONTAINMENT PRESSURE AND TEMF'ERA'>>RE RESPONSE MS'SLOT) WITH LOSS OF ONE CONTAINMENT COOLING TRAIN
~
(SHEET 3 OF 5)
Figure 6.2.1-6
O 25% POWER VAPOR TEMPERATURE VAPOR TMA
~ 400.7 F 0 80 SEC 400 4P U
L'>>
IZ
> 20 z
Iz,0'J 10 SUMo TEMPERATURE
~PRES JRE SUMP T 257.1 F
MA g 380 SEC MAX 0.5 PSIG
$ 160 SEC I SPRAY TRAIN ON e 80 SECONDS SUMP
>>C a.
C C
2Q I
100 I
I I
I III I
I I
I III 100 TIME FOLLOWING BREAK ISECONDSI I
I III 1000 I
I I Ill 10000 R~P~~
Mi~
4~~IkED 7 RC Fe SEp F lc u~
Palo Verde Nuclear Generating Station FSAR CONTAINMENT PRESSURE AND TEMPERATURE
RESPONSE
MSL3 (SLOT) WITH LOSS OP ONE CONTAINMENT COOLING TRAIN tSheet 4 o~ 5)
Fj.gee 6 ~ 2 ~ 1
I 0
'll
~ I I I I
I:
I
~
I f
I II I
t I
I I I 0'
0 L
,P
~ ~
~
~
S
~
~
~
9 0% POWER VAPOR TEMPERATURE VAPOR TMAX 39 F
6 80 SEC 403 UMP Th'AX 256.6 F 8 342 SEC I
30=
40 Q
M C
30 CL I
10 SUhIP TEhIPERATURE RESSU
'i RE I SPRAY TRAIN ON G 80 SECONDS PMAX Q 194 SEC 20 10 I
I I
I I III 10 100
(
I I
I I
I l 1000 I
1 I
10000 TIME FOLLOWING SREAK ISECDNDS)
+~LA~ Mt~ L~MHED
'P~Pt-~
FlCi&RE Palo Verde Nuclear Generating Station FSAR CONTAINMENT PRESSURE AND TEMPERATURE RESPONSE MS'SLOT)
WITH LOSS OF ONE CONTAZNMENT COOLING TRAZN (Sheet 5 of 5)
Figure 6.2.1-6
CC0
&POWER TMAX 393. 1fF
@84 SEC ~
MAX 0 180.5 SEC 52 360 320 c:
28C I
2CC MAX 264,7F C'40 SEC 42 0
CL I
, R C
32 g
200 1 SPRAY TRAIN ON 0 84 SEC 27 160 22 120 100 TIME FOLLOWING BREAK (SECONOS)
I
~
I I
1000 ss 17 1000G PROPOSED Palo Verde Nuclear Generatinp Station FSAR CONTAINYENT PRESSURE AND TEMPERATURE
RESPONSE
HSLEI (SLOT) WITH LOSS OF ONE CONTAINMENT COOLING TRAIN (SHEET 5 OF 5)
Figure i6.2. 1-6 AC
~ 293 BTU(HR ~ FT2 oF
-MAX 0 15 SEC.
DOUBLE END DISCHARGE LEG SLOT B c AK MAXIMUM CCS 250 z
O 200 0
~ C cn m 2
LL tc c 150 I
I C I-ill Cl C
Eh 100 z
0z0U 50 TRAN:ON REG N
RECIRCULATION BEGINS C~
2050 SEC.
TAGAM<
FUNCTION I
I I >>
UCHIDA F UNCTION I
I I
I >>>>
I I
I I>>I I 11>>>>
i
> ~1>>f!
10 102 10 10 105 v 10 TIME FOLLOWING BREAK (SECONDSI P~~AE,D F1G. L)RE Palo Verde Nuclear Generating Station FSAR CONDENSING ~T-TRANSFER COEFF ICIENTs DOUBLE-ENDED
~
DISCHARGE LEG SLOT BREAK, MAXINJM ECCS Il.
~
4 ~ova C
0
R 320 DOUBLE ENDED DISCHARGE LEG SLOT BREAK MAXIMUMECCS 240 I
200 EN RO RIG tH ZC EE 1BP l
120 C)
~ea So TAGAMI FUNCTION TRANSITION REGION UGHIOR FUNCTION 1 SPRAY TRAIN ON C'91 SEC
~ RECIRCULATION C-'117 SEC 0
1
.....10.
'I I I II I
I I I I I
I I
I I
I I
I IIIIEI t
10
'10~
02
-103 10 4
~
~
TIME FOLLOWING BREAK (SECONDS)
PROPOSED Palo Verde Nuclear Generating Station FSAR CONDENSING HEAT-TRANSFER COEFFICIENT R DOUBLE-ENDED DISCHARGE IZC SLOT BREAKR MAXINJH ECCS Figure 6.2 '>>7
'l
~
l 1
120 IL 0
CVIIL Z
100 I
CI Iz 80 O
0 CJ 60 z
l 4~
I z
20 Z
0Z0 0
I t
I I
I 102> POWER 1 SPRAY TRAIN ON C BO SECONDS I
1 I
I I III 10 100 TIME FOLLOWING BREAK (SECONDS) 1000 10000 p',ppt ~
wlvA. g~kQ~O P'j+DP~g~
- PIQOR.Q Palo Verde Nuclear Generating Station
. FSAR CONDENSING HEAT-TRANSFER COEFFICIENT (UCHZDA FUNCTION)
MAIN STEAM LINE SLOT BREAKS WITH LOSS OF ONE CONTAINMENT COOLING TRAIN Figure 6.2.1-9
160 140 R
120 C)
CJ LM ~ 100 C'h~
CZ Ik I
80 60 a
CD CJ 40
'02~8 POWER~
~ O'Yi POWER 1 SPRAY TRAIN ON 0 84 SEC 20 10 100 I
~
I I
I I
I 1000
'000 TIME FOLLOWING BREAK (SEeoNOSI PROPOSED Palo Verde Nuckar Generatj~
S)a<jog'SAR CONDENSING ~+ T-TRANSFER COEFFICIENT (UCHIDA FUNCTION)
MAIN STEAM LINE SLOT BREAKS bITH LOSS OF ONE CONTAIN>ZNT COOLING TRAIN Figure E
2 I-9
0 r,4
10 0
10 SUMP WATER RE CI RCULATIO BEGINS C'05 SEC I2 H.P. SI PU
. SI p,'I Se+A <FE ce P
I 0C 10 C
,c~~-
1 RAY RAIN ON 0 87 SEC R
ACTOR VESSEL WATER 40 10 I
I IIIII 10 I
I I I III 10
'c~
REACTOR A
VESSE L CALCULATIONS IN ITIATED 0 500 SEC.
I I
I I II I
10 UX 4I I
C
(
C I
I I I II 10 I
I I
I I II 10 I
I I
II 10 TIME FOLLOWING BREAK (SECONDS)
Pmpose~
FlCu~
I Palo Verde Nuclear Generating Station
~ FSAR CONTAINMENT ENERGY INVENTORY~
DOUBLE-ENDED DISCHARGE LEG SLOT BREAK, MAXIMUMECCS Fi re 6.2.1-10
0
11 1010 10 VAPOR 108 SuI'P EACTOR VESSEL VcA 1O'O' SPRAY~
TRAIN ON 0 91 SEC REACTOR VESSEL CALCULATION IN IT IATE 0 Cl 500 SEC RECIRCULATION C 2117 SEC 1O'O'0 10 10 10 10 TIMF. FOLLOWING BREAK ISECONDS)
'lo l
l l I I I I
l l
I I I I I I
I i i(I I
I I
I I I I I I
l I
~
I I I I PROPOSED Palo Verde Nuckar Generating Station FSAR
'ONTAINMENT ENERGY INVENTORY, DOUBLE-ENDED DISCHARGE LEG SLOT BREAK,
~IMUM ECCS Figure 6.2.1-10
1O' 102$ c POWER CONTAINhlENT.
ATh'.OSP HE R E 1O'TRUCTURAL HEAT SINKS CONTAINh!cNT S'Jh',P 1O'0 I
I I
I I
10 1 SPRAY TRAlh ON 0:
80 SEC I
I I
I I
I I
I I
I III 1000 I
I I
I I III 1000 TIME FOLLOWING BREAK ISECONDSI gmWc~
Vkop Palo Verde Nuclear Generating Station FSAR CONTAINHEhT ENERGY DISTRIBUTION HSLB (SLOT)
WITH LOSS OF ONE CONTAINHENT COOLING TRAIN
{Sheet 1 of 2) pgcpxye 6.2.1-1.2
0 102K tOWER 10'0 106 1 SPRAY TRAIN ON CI 84 SEC 10 10 I
I I
I I
101 102 I
I I
I I
10 I
I I
10 TIME FOLLOWING BREAK (SECONOS)
PROPOSED Pdo Verde Nuckar Generating Station
FSAR CONTAINMENT ENERGY DISTRIBUTION MSLB (SLOT) WITH LOSS OF ONE CONTAINMENT COOLING TRAIN (SHEET 1 OF 2)
Figure 6.2.1-12
10
@ 0% POWER CONTAINMENT ATt~lOS+HERE 10 10 STRUCTURAL HE<<T SINKS CON
~ AINMENT SUte'sP go O~
C A
10 I
I I II 1 SPRAY TRAIN~
ON C-'
80 SEC I
I I
I I
I I
II 10 1000 TIME FOLLOWING BREAK (SECONDSI 1003:
~-TT<c.Hew PRC MS~
PIC u~
Palo Verde Nuclear Generating Station FSAR CONTAINMENT ENERGY DISTRIBUTION MSLB (SLOT)
WITH LOSS OF ONE CONTAINYE¹ COOLING TRAIN
{Sheet 2 of 2)
Figure 6.2.1-12
1 0
'I 0
10 Q 0N POWER 108 u~>'"
1 SPRAY TRAIN ON C 84 SEC 10 10 I
I I
I I
I 10 I
I I
I I
I I I
I I
I I III 10 102 TIME FOLLOWING BREAK (SECONDS)
I I
I I IIII 10
n gC PROPOSED Palo Verde Nuclear Generating Station FSAR'ONTAINMENT ENERGY DISTRIBUTION MSLB (SLOT) WITH LOSS OF ONE CONTAINMENT'.
COOLING TRAIN (SHEET 2 OF 2
)
Figure 6.2.1-12
7
450 POSTS 5LB IN CQNTAI Ih1ENT eoulohj'BNT OUALIFICATIQNTfh1PER g URE (370 F) e3 350 0
3PA I
255 O
200 I
I I
I I
The
368.5 F
PMAX = 45 PSI G 50 I-c2 Z
40 2
'C I.
30 c
fl 0:
c:
20
~150 co~
<e 1 SPRAY TRAIN I ON C-'p SSCONOS 10 10 100 TIME FOLLOIVINGBREAK ISECONDSI 1000 IG.(hS
~
~
~
Palo Verde Nuc)ear Generating Station FSAR EQUIPMENT QUALIFICATION PRESSURE AND TEMPERATURE
RESPONSE
MSLB (SLOT) AT
, 102l POWER Figure 6.2.1-20 Febnsirv 1cIRC hesaA a
~ s
P
440 e tm ~OWER MAX
@168.5 SEC 57 52 360 320 Uo
~ 280 CC 240 200 e
TMAXc 367.8$ F C 84 SEC 1 SP RAY TRAIN ON MAX
=
C 310 SEC 1 SPRAY TRAIN ON 84 SEC 47 0
C 37 I
R'2 C
27 160 22 120 I
I I
I I
~
~
10 I
I
~
~
100 1000 TIME FOLLOWING BREAK (SECONDS) 17 I
I I
I I
~
~
10000 PROPOSED Palo Verde Nuclear Ceneratinp Station FSAR EQUIPMENT QUALIFICATION PRESSURE AND TEMPERATURE RESPONSE MSLB (SLOT)
AT 1021 POWER Figure 6.2.1-20
0
PANGS F SAR EMERGENCY CORE COOLING SYSTEM 7.
The containment spray system is designed and constructed in conformance with ASME III Class 2
requirements.
The material used in this system is austenitic stainless steel type 316 or 304 or other C-E approved compatible material.
System/Component Arrangement 2.
3.
The piping arrangement is such that the maximum and minimum head loss requirements presented in CESSAR Table H~~are met.
Q,&.2. -5.'o The piping for each safeguards train is designed such that the,top of the piping junction.of the pipe runs to the refueling water tank and the containment recirculation sump is located at least 16 feet below the top of the recirculation containment
sump, which is 4 feet below the minimum
'ater level in the containment during recircula-tion.
The containment minimum pressure below the refueling water ambient pressure does not exceed 3 psi as described in section 6.2.1; therefore, an increase in the differential height of the piping junction for containment underpressure in excess of 3 psid is not required.
Frictional losses in the safeguard pump suction piping between the containment sump and the junction with the RVZ are less than 5 feet.
The location and elevation of the HPSI pumps are indicated on figure 1.2-4.
The HPSI pumps are
.located in the auxiliary building as close as practical to the containment.
6-3-21'
~
I
PIGS PSAR EMERGENCY CORE COOLING SYSTEM a ~
b.
The elevation of the HPSI pumps is such that the available NPSH is at least 22 feet during the recirculation mode when the pumps take suction from the containment sump.
'In determining this elevation, no credit was taken for subcooled water in the containment sump following a LOCA.'he available NPSH was calculated at t.he pump suction that was assumed to be a
maximum of 2.5 feet above its foundation elevation.
C.
d.,
The available NPSH was calculated taking into consideration the concurrent operation of the low pressure safety-injection and containment spray pumps tabulated in CESSAR
bl G.5.2.-5. o Credit was not taken for'ater trapped above the containment floor.
The requirements on location and elevation applied to the HPSI are applied-to the location and elevation of the LPSI pumps except. that the pump's suction is 14 inches below its supports.
The elevation. of the LPSI pumps is such that the minimum required NPSH of 22 feet during injection at a runout flowrate of 5100 gal/min per pump is satisfied.
The design calculation assumes:
a.
Concurrent HPSI,LPSI, and containment spray pump operation and flow in conform-
- ance with.CESSAR Table 6.3-5a.
b.
A refueling water tank temperature of 100K.
6. 3-22 4S
~
~
S to Kl'b
O.

v t e Letter Sequence Draft Other
TAC:60706, (Open)
Initiation Request Acceptance... Administration Meeting Questions Answers Results Approval... Other: ML17299A490, ML17299B011, ML17299B110
MONTHYEAR1985-08-12 [Table View]

ML17299A717

Text

SUBJECT:

Subject:

Reference:

Dear Mr. Knighton:

Response

Subject:

Reference:

Subject:

Dear Hr. Knighton:

SUMMARY

RESPONSE

RESPONSE

RESPONSE

RESPONSE

RESPONSE

RESPONSE

Navigation menu

Search