ML17297A885
| ML17297A885 | |
| Person / Time | |
|---|---|
| Site: | Palo Verde  |
| Issue date: | 09/28/1981 |
| From: | Van Brunt E ARIZONA PUBLIC SERVICE CO. (FORMERLY ARIZONA NUCLEAR |
| To: | Tedesco R Office of Nuclear Reactor Regulation |
| References | |
| ANPP-19015-JMA, NUDOCS 8109300362 | |
| Download: ML17297A885 (130) | |
Text
REGULA, ORY INFORMATION DISTRIBUTI I SYSTE'>>I (RIDS)
ACCESSIOV>> NBR: 8109300362! DOC ~ DATE: 81/09/28 NOTARIZED NO ~ -
DOC iT FACILi:STN 50-528 Palo Verde Nuclear Station< Unit 1E Arizona Publi 05 STN 50-529 Palo Verde Nuclear Stat'ionP Unit 2g Arizona Publi 05000529 STiV-50-530 Palp Verde Nuclear Station~ Unit 3r Arizona Publi 05000530 AUTH!, NAME! AUTHOR AFFILIAT'ION VAN BRUVITP EI,E>>. Arizona Public'ervice Co.
REC IP ~ VAMEI REC'IPZ ENT AF F IL>>I ATION TEDESCOER ~ Lrs Assistant Director for L.icensing
SUBJECT:
Forwards revised responses to VRC Questions 440 1 tnrough
'7 F
440 'esponses will be'ncorporated into future FSARr amend, DIsTRIBUTIGN cDDEI: BODis cDPIEB REcEEUED:LiTR L'NGL< s?zE'.: j>>r TiITLEI PSAR>>/F SAR AMDTS and Re at.ed Corr espondence 1
NOTES:Standar dized Plant. 1 cy:C'r imes 05000528 Standardized P l ant ~ 1 cy.' Grimes OS000529 Standardized Plant. 1 cy'.C" Grimes 05000530 RECIPIENT COPIES RECIPIENT COPIES I O- CODE/I>>IAI~IEI LTTR ENCL ID CODE/NAME LTTR ENCI.
ACTiION: A/D>> Lil CEiVSNG 1 0 LIIC BR 43 BC 1 0 LIICI BR 03 lA 1 0 KERRIGANtJ ~ 04 1 1 INTERN ALi: ACC:ID EiVAL 1 ~
1 AUX SYS BR 27 1 1 09 BR26'HEM EVG BR 11 1 1 CONT SYS BR 1 1 CORE PERF BR 10, 1 EFF TR SYS BR12 1 1 EQUIP QUAL BR>>1 3i 3 3 GEOSCIENCES 28 2 2 HUM FACT ENG 40. 1 1 HYD/GEO BR 30 2 2 III Ci SYS BR 16" 1 ICE 06 3 3 IE/EPDBI 35 1 1 IE'/EPLB 36 3 3 L>>IC'UID BR 33>> 1 1 LIIC QUAl'R 32 1 1 MA'I IE VG BR 17 1 i>>IECH ENG BR 18 1 1
>>>>IP A 1 0 OELD 1 0 OP LIC BR 34 1 1 PO'HER SYS BR 19 1 1.
PROC/TST REV 20. 1 1 QA BR 21 1 1 RADi E S BR!22>> REAC SYS BR 2'3 FILE,'1 A
R2)i 1
1 1
1 SIT ANAL BR 24 1
1 1
1 ST >>VG B 1 1 EXTERNAL(: ACRS 41>> 16 16 FEMA REP DIV 39 1 1 LPDR 03i 1 1 IVRC PDR 02'T'I 1 1 NSZC 05>> 1 S 1 1>>
gt;y g2 l1).
TOTAL IVV'>>IBER OF COP IE S REQUIRED: LiT TR 6 ENCL.
4 D O IRUilKItal 48ltlIHPHtgGS tWQHHW STA. P.o. SOX 21666 - PHOENIX, ARIZONA 85036 September 28, 1981 ANPP-19015 JMA/TFQ Mr. R. L. Tedesco Assistant Director for Licensing Division of Licensing gg Office of Nuclear Reactor Regulation SPP2$ III U. S. Nuclear Regulatory Commission 4wll~l Washington, D.C. 2055 Subj ect: Palo Verde Nuclear Generating Station (PVNGS) Units 1, 2 and 3 Docket Nos. STN-50-528/529/530 File: 81-056-026 G.l.10
References:
(A) Letter from R. L.-Tedesco, NRC, to E. E. Van Brunt, Jr.,
dated June 22, 1981, subject: Request for Additional Information PVNGS (RSB)
(B) Letter from EEVB to R. L. Tedesco, ANPP-18786, dated August 28, 1981 (C) Letter from EEVB to R. L. Tedesco, ANPP-18881, dated September 9, 1981
Dear Mr. Tedesco:
. al-On September 10, 1981, NRC/RSB reviewers met with APS represelItatives to discuss reference (C), which were outstanding items from a similar meeting on September 4, 1981. These meetings discussed reference (B),
which responses to the staff's request for information, reference (A).
Attached are revised responses to NRC Questions 440.1 through 440.87, for your use. These responses will be incorporated into the FSAR in an upcoming amendment.
Please contact me if you have any further questions on these matters.
Very truly y u s, Mu APS Vice President, Nuclear Projects ANPP Project Director EEVBJr/TFQ/kj'<09300362 8i 09+8+ /J ggs PDR MOCK 05000528,
'PDR'c:
J. Kerrigan (w/a) P. L. Hourihan (w/a) o C. Liang (RSB) (w/a) A. G. Gehr (w/a)
F R
}e
STATE OF ARIZONA )
) ss.
COUNTY OF MARICOPA)
I,'dwin E. Van Brunt, Jr., represent'hat I am Vice President Nuclear Projects of Arizona Public Service Company', that the foregoing document has been signed by me on behalf of Arizona'Public Service Company with full authority so to do, that I have reap such document and know its contents, and that to the .best of my knowledge and belief, the statements made therein are true.
Edwin E.Van Brunt, Jr.','i'-'worn to before me this ay of ~ 1981.
(g No ary Public My Commission expires:
Mc AP 3
w,">.
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APPENDIX 5A
~y
'y r<<
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..s Il ' " ( Q""""" (5.2.2)
A description of the design features which will .be used to mitigate the consequences of overpressurization events while operating at low temperatures is not provided in the CESSAR System 80 FSAR. Provide a description of the features which will be provided on the CESSAR System 80. Specific design criteria regarding overpressurization protection while operating at low temperatures are as follows:
- l. 0 erator Action: No credit can be taken for operator action for'10 minutes after the operator is aware of a transient-h y.
.2. Sin le Failure: The system must be designed to relieve the.
pressure transient given a single,gilur'e in addition'to the '<<<<y-failure that initiated the pressure transient.
~y'y': '*y y y My y basis consistent with the system's employment.
- 4. Seismic and IEEE 279 Criteria: 'deally, the system should meet seismic Category 1 and IEEE 279 criteria. The basic objective is that the system should not be vulnerable to a common failure that would both in'itiate a pressure. transient
- t l ~
September ..19 81 =;". " .;.:.;... Amendment 6
'7-28-81.
PVNGS FSAR APPENDIX SA and disable the overpressure mitigating system. Such events as loss of instrument air and loss of offsite power must be considered.
An alarm must be provided to monitor the position of the pres-surizer relief. valve isolation valves to assure that the over-pressure mitigating system is properly aligned for shutdown conditions.
In demonstrating that the mitigation system meets these criteria, the applicant should include the following information in his submittal:
- 1. Identify and justify the most limiting pressure transients caused by mass input and heat input.
- 2. Show that overpressure protection is provided (do not violate Appendix G limits) over the range of conditions applicable to shutdown/heatup operation.
- 3. Identify and justify that the equipment will meet pertinent parameters assumed in the analyses (e.g., valve opening times, signal delay, valve capacity) .
~ +wW4
- 4. Provide a description of the system including relevant P&I drawings.
- 5. Discuss how the system meets the criteria.
- 6. Discuss all administrative controls required to implement the protection system.
RESPONSE: The response will be provided on the CESSAR docket.
g 5 Amendment 6 5A-2 September 1981 07-28-81
PVNGS FSAR APPENDIX 5A Q '" ". ( Q"""" .)P (5.2.2)
Provide details of your proposed pqeoperational and in>tial startup test program to show that they are consistent with the requirements of Regulatory Guide 1.68,
~Tie. Q<~ioQQ~ S bs o~(iQicd o~~CaSSW doCkd-4r RESPONSE- ~ .
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Q * ( Q"" (5.2.2)
Check valves in the discharge side of the high pressure safety injection, low pressure safety injection, RHR,. and charging systems perform an isolation function. in that they protect low pressure systems from full reactor pressure. The staff will require that these check valves be classified ASME IWV-2000 Category AC, with the leak testing for this class of valve being performed to code specifications. It should be noted that a testing program which simply draws a suction on the low pressure side of the outermost check valves will not be accept-able. This only verifies that one of the series check valves is fulfilling an isolation function. The necessary frequency Mill be that specified in the ASME Code,,except in cases where only one or two check valves separate high to low pressure systems. In these cases, leak': testing will be performed at rQ;Q:i~>&'Al~+iM-"-
each refueling after the valves haye+~yeiY=exercised;" Identify
~
all check valves which should be classified Category AC as per the position discussed above. Verify that you have the neces-sary test lines to leak test, each valve. Provide the leak detection criteria that will be in the Technical Specifications.
cheat, RESPONSE: The following valves are classified Category AC A
as described above:
Safety Injection (SI) Valves V-p215J 217'25r227 541,542 and 543 tuiI( 'Q gravid on We L~ARdoctet Qr .c4~&
.. 235 237s245s247J 540, 5fQ> Q3g
~~
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J@g~ Qip<e(Q September. 198l-, .- ... ., ".. 5A-.3 0.7-28-81" - ..'
.',.:, - Amendment 6
, aS ShaOh ih Figure, &.3 -I>
Adequate test connections and linesphave been provided to facilitate testing of the above listed valves. to ASME IWV-2000 Category AC requiz'ements.. The leak detection c4 I Bill~,~
criteriazwzll be included in the Technical Specifications ~
e Q (5A)
On page 5A-2, it ie,igd~ated A All'A that a negative Doppler coeffici- '
ent of -~8 x 10 phd~
sure transient (loss of load).
is assumed in the bounding overpres-It is our position that over-pressure protection of system be demonstrated without taken credit for either doppler or moderator temperature reactivity feedback (SRP 5.2.2,Section III.6). Reanalyze the bounding overpressure transient without credit for doppler feedback, demonstrating that primary system pressure does not exceed 110%
of the design pressure.
RESPONSE: The response will be provided on the CESSAR docket.
OUESTION 5A.7 (NRC Question 440.5) + (5A)
On page 5A-1, it is indicated that the worst case transient, loss of load, in conjunction with a delayed reactor trip, is the design basis for the primary safety valves. It is our 0'
position that the high pressure reactor trip or second safety
~
grade trip signal, whichever occurs later, should be used for sizing the primary system safety valves. Confirm that the CESSAR System 80 safety valves are sized sufficiently to accom-modate a reactor trip on the second safety grade trip signal.
RESPONSE: The response will be provided on the CESSAR docket.
Amendment 6, SA-4 September 1981 07>>28-81
Palo Verde must have the capability to take the plant from full power to a cold shutdown using only safety grade equipment, per the requirements of BTP RSB 5-1. Address your compliance with all provisions of that position and respond to the detailed question below.
Question Describe the sequence for achieving a cold shutdown condition within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />, assuming the most limiting single failure with only onsite power availability. Identify all man-ual actions inside or outside containment that must be performed and discuss the capa-bility of remaining at hot standby until manual actions (or repairs) can be performed.
Ouestion If the steam generator dump valves, operators, air and power supplies are not safety grade, justify how you would cool down the primary system in the event of loss of offsite power and an SSE.
Question Describe the sequence for depressurizing the primary system using only safety-grade systems, assuming a single failure. Identify all manual actions inside or outside containment that must be performed.
Question 'l C.. Discuss the boration capability using only safety-grade systems, assuming a single fail-ure. Identify all manual actions inside or outside containment that must be performed.
If the proposed boration method utilizes the charging pumps (assuming a letdown line failure is proposed), provide an evaluation September 1981 ,5A.8-1 Amendment 6 I
07-30-81
of this approach with regard to concentration of boron source and liquid volume in primary system.
Question 2. Discuss the provisions for collection and containment of RHR pressure relief valves discharge.
Question 3. Describe tests which will demonstrate ade-quate mixing of the added borated water and cooldown under natural circulation conditions with and without a single failure of a steam generator atmospheric dump valve. Specific procedures for plant cooldown under natural circulation conditions must be available to the operator. Summarize these procedures.
Question 4 ~ Discuss the availability of the Seismic Category I auxiliary feedwater supply for at least 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> at, hot shutdown plus cooldown to the RHR system cut-in based on longest time for the availability of only onsite or only offsite power and assuming a single failure. If this cannot be achieved, discuss the availability of an adequate alternate Seismic Category .I water source.
Question 5. What provisions in'atural circulation cool-down methods have been made to account for possible upper head void formation?
'RESPONSE: iiiwill v espouse.
be provided on the CESSAR docket. AddHionel c)aright c+(on is PevldecI o.s Qllou)~.
Amendment .6 5A.8-2 September 1981 07-30-81
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The Natural Circulation Boron Mixing Test performed at SONGS will be reviewed for applicability to PVNGS by the plant staff, and a post-test report will be submitted to the NRC upon completion of that review.
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PVNGS PSAR APPENDIX 5A IK *' " ("" Q""" * ~
)v (5.4.7)
.>royide;-gptailed information on the sizing criteria used to
-de~tprminq the relief capacity of the SDCS suction line pressure relief valves.
Did the .version of the ASNE code that the SDCS relief valves pyrt sided to require establishing liqui.'d or two-phase relief sagacity wth testing? If so, describe in detail the test
-;;.'-. -,program;aqdz results. If the 'liquid or two-phase relief capac-
~
., -,<<igy was ~not established by test, show that the difference Pe)ween +e rated and maximum required relief capacity is more
-:c.,than sufficient to bound liquid and two-phase relief rate
'certainties.
e s.&. Prpvide details on the alarms and indications which would inform thg,eyerators that a SDC suction line isolation valve
>yq;c-osege"bile. the plan" is in shutdo z cooling. Is there
,v p;,>
ya aug .common,qfailure. which would result in both valves being closed while in shutdown cooling.
'pep'..LPSI preamp mini flow isolation valves are closed during shytgown cooking, what would prevent pump damage if a pressure gypsienM wede to occur which caused RCS pressure to exceed LPSY deadhead pressure.
- -";>C Whenythe pla~ilis in the SDCS mode, is there any single failure
~o, whjgh could cause the suction of both SDC pumps to be switched frog the hot leg piping to'he dry sumps?
RESPONSE: <<
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.i,",. lg adk hen, 'i'period:s'~ppliel b~ indg<<de& ~<<- s~/~plies are
'(.<'u gvidcd to "Iso(eke coo(i<<<<q sysieun M i'm puss<<re parian ~$ ~ s'ttukdm<<<<
September 1981 Amendment 6
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- PANGS F SAR APPENDIX 5A OUESTION 5A.10 (NRC Question 440.8} P (5.4.7)
Provide the following information related to pipe breaks or leaks in high or moderate energy lines outside containment associated with the RHR system when the plant is in a shutdown cooling mode:
- 1. Determine the maximum discharge rite from a pipe break in the systems outside containment used to maintain core cooling.
- 2. Determine the time available for recovery based on these discharge rates and their effect on core cooling.
- 3. Describe the alarms available to alert the operator to the event, the recovery procedures to be utilized by the operator, and the time available for operator action.
A single failure criterion consistent with Standard Review Plan 3.6.1 and Branch Technical Position APCSB 3-1 should be applied in the evaluation of the recovery procedures utilized.
RESPONSE
Amendment 6 5A-6 September 1981 P7-28-81
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(5.4.7)
Indicate whether there are any systems or components needed for shutdown cooling which are de-energized or have power locked out during plant operation. If so, indicate what actions have to be taken to restore operability to the components or systems.
It is the staff's position that all operator actions necessary to take the plant from normal operation to cold shutdown '(SDCS entry) should be performed from the control room. If the present design does not meet this position, please commit to revise e, res it ~
accordingly.
se <II( be.divided
RESPONSE
the CESSAR docket.
September 1981 5A.11-1 Amendment 6 07-30-81
~,' ~ ~ L UESTION 5A.12 (NRC Question 440.10) M (5.4.7)
Provide additional information regarding the power sources supplied to the SDCS isolation valves. The staff's that a single failure of a power supply will not prevent position is tion o the SDCS when RCS pressure ezceeds its design pressure. isola-
~
Additionally loss of a single power supply cannot result in the
. inability to initiate at least one 100 percent shutdown cooling train.
RESPONSE
vie.'es~~ ~pit g Prnvrcfccf on4@C~QRdocke+
dko~algl, The power supply to the shutdown cooling isolation valves is given below. Valve arrangement is shown on figure 6.3-1.
Train A: UV-655 (Note 1)
UV-653 (Note 3) 1 UV-651 (Note 1)
NOTES:
Train B:
UV-656 (Note 2)
UV-654 (Note 4)
UV-652 (Note 2) (5A-7
~ ~
j I
I
~
I
~
- 2. Fed
~<'kW from>Cla~s IE
..:-:.'.'CC
~
E-PHB-M36 I'.
Fed from Class IE Channel C battery through Class IE Inverter E-PKC-N43
~ ~
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~SPi 8 tt> c 5/i-3'. 444 lO ac power Supplies indicafccf provide/t iver 4 Au'. valve. it/era]ar (cuir).
Traitt A valves ave. cctttplek+ I'ndegndctvt- Wtn Mtti n
~tg Et, eac4 a rcdatdattf Parti/A'/
valves, Shuhb>uiu COOl(
lu fcrrcs .
pa ~
viewing
&LCA T4'l-ckre, tto tnuri COnrrS> Ob Sing/e ctr contntort
~B ValteS i itre catt csu in /oss o$ 56iihkuitt coo/in) cApak/i~/~
ei4kcr duc + m &ilare oP a Iralvc.& opet or Qi/rcte-.
vitlve to cfos<. 7Aus 7Yhlcts ttteeks Qi lure Crikcnlsn ~
~ slhg/K p>thieved 4atitt 2.seI arcctr .dtctnnela..
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PVNGS FSAR QUESTION 6A.28 (NRC Question 440.11) M (6.3)
Discuss the provisions and precautions for assuring proper system filling and venting of ECCS to minimize the potential for water hammer and air binding. Address piping and pump casing venting provisions and surveillance frequencies.
RESPONSE: The safety injection piping will be maintained filled with water. This will minimize the potential for water hammer. All piping is provided with high point vents and low point drains. The centrifugal pumps are vented through their discharge pipes. The pumps use a casing drain for draining. The containment spray headers will be maintained full up to elevation 115 feet. The safety injection pumps will be tested monthly to satisfy the requirements of ASME XI. To assure a full system, procedures will be developed ensuring proper inspection of key points a
.6aM ~~ ~ 3/~ 3 September 1981 '-'"".:
~ / I I 6A. 28-1 Amendment 6 07-30-81 I
pp ~ I ~
~ .
PVNGS F SAR UESTION 6A.29 (NRC Question 440.12) (6.3.3)
Section 6.3;3.2.2 states that the worst single failure for the large break LOCA is the failure of one of the low pressure.
pumps to start which will result'in a minimum amount of safety injection water available to the core. Explain why the single failure of a diesel generator, which results in loss of one HPSI train and one LPSI train, is not the worst single failure for the large break LOCA with respect to the amount of safety injection water available to the core in post LOCA operation.
RESPONSE: The response will be provided on the CESSAR docket.
s PO'cK Ic ~
September 1981 6A.29-1 Amendment 6 07-30-81
PVNGS FSAR QUESTION 6A.30 (NRC Question 440.13) ~ (6.3)
Identify'll ECCS valves that are required to have power locked out and confirm they are included under the appropriate Technical Specifications, with surveillance requirements listed.
,RESPONSE: The response will be provided on the CESSAR docket.
Q lb+4 P w 1P P,t September 1981
' . '. \
6A.30-1 D7-30 81 I
Amendment 6
Consideration should be given to the possibility that local manual valves (handwheel), could go undetected in the wrong position until a postulated accident occurs. Appropriate administrative controls or valve position indication are examples of methods to be considered to minimize this possi-bility. Provide a list of .all critical manual valves and address the actions that will be implemented to assure all critical valves are properly positioned.
Identify all manual valves which have locking provisions.
It is our position that limit switches which enable valve position to be indicated in the control room should be installed on all manually operated and normally locked ECCS valves.
In addition a recent event (Docket 50-320, LER 78-20/3L, 4/21/78) has brought to our attention that the automatic operation of some motor operated valves can be disabled when the manual handwheel pins are engaged. Identify all critical motor operated valves associated with the CESSAR 80 design that have this design eature and describe the controls and procedures utilized to prevent the inadvertent disablement of the automatic operation of these valves.
RESPONSE
The vespowse ~ill be provided o~ Se CERAM doeke&.
September 1981 "
6A.31-1 Amendment 6
(! * . ( Q (6.3)
Identify the plant operating conditions under which certain automatic safety injection signals are blocked to preclude unwanted actuation of these systems. Describe the alarms available to alert the operator to a failure in the primary or secondary system during this phase of operation and the time available to mitigate the consequences of such an accident.
RESPONSE
The response will be provided on the CESSAR docket.
(fAg,
'p September 1981" ".-: .'.':" "-:::.":;'"6A.32-1 .:;:...:.," A end ent 6
PVNGS FSAR ll '*' " ( I!"" "" (6.3)
The information in the CESSAR 80 FSAR regarding post-LOCA passive failures is not complete. It is the Reactor Systems Branch position that detection and alarms be provided to alert the operator to passive ECCS failures during long-term cooling which allow sufficient time to identify and isolate the faulted ECCS line. The leak detection system should meet the following requirements:
Identification and justification of maximum leak rate should be provided.
- 2. Maximum allowable time'or operator action should be provided and justified.
- 3. Demonstration should be provided that the leak detection system will be sensitive enough to initiate (by alarm) operator action, permit identification of the faulted line, and isolation of the line prior to the leak creating undesirable consequences such as flooding of redundant equipment or excessive radioactive fluid. The minimum time to be considered is 30 minutes.
It should be shown that the leak detection system can identify the faulted ECCS. train and that the leak is isolatable.
- 5. The leak detection system must meet the following standards:
- a. Control Room Alarm
- b. IEEE 279-1971, except single failure requirements
RESPONSE
t gNkc
- 3. The level instrumentation is mounted in each safety injection pump room sump. This provides a high water level alarm in the control room after an accumulation of 3.5 gallons of water in the sump. Each safety September 1981 Amendment 6 l6
~
07-30-81 q ~
L
~ ~
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~ O'FSX
+0 -O . @Ver 4P y'o'-o" M~,~
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i'
~. lb injection pump room sump high water level alarm is a IE annunciation in the control room. This level is sufficient to provide isolation of the leak by appropriate operator action within 30 minutes. =This action will consist in part of shutting suitable isolation valves to stop the leak. This action will also include steps to isolate the leaking train.
- 4. The safety injection leak detection system consists of individual level switches in each train pump room.
Individual control room IE annunciation windows enable identification of the leaking train. See 6A.18.3 for leak isolation methods.
- 5. The safety injection leak detection system consists of a IE (safety grade) switch in each pump room for each train of the:
High Pressure Safety Injection Pump Low Pressure Safety Injection Pump Containment Spray Pump Each level switch actuates a IE annunciation in the contro room. The train "A" pump wooms are monitored by channel "A"
'TkaA A instrumentation powered by Class IE power. The train "B" pump room are onitored by channel "B" instrumentation powered b ass IE power. The system complies,with IEEE Standard 279-1971 except for single failure requirements.
FN@ '4/(dig eKkauk Factlak'bw wLovl/gaol-g iNiiii {oy ggglg aliis vet~<<s' areas Subg~d + )~kg< M~
~ />-3'- s@6 esMikal RLI-/i/-5 4/~hb~
CSF y'Qc~'n~l+'by u~ifS +4&S~ are'. cfarcr:beg
qq+~ ii~ 7/piQiI. gogiQw senslk vi/fes 5gc >n fl g and cz~ Wg~ ~ p~vi~ 8Qr(y de~M o4-r.ec.irc~(~flog bop f~kge.,
Amendment 6 6A.33-2 September 1981 07-30-81
The acceptance criteria in the Standard Review Plan for Section 6.3 states the ECCS should retain its capability to cool the core in the event of a single active or passive failure during the long-term recirculation cooling phase following an accident. Demonstrate that CESSAR 80 ECCS design has this capability.
RESPONSE: The response will be provided on the CESSAR docket.
September 1981 '.-'""'"',:: 6A.34-1 Amendment 6 07-30-81
APPENDIX 15A OUESTION 15A.12 (NRC Question 440.40)~ (15.0)
As part of the CESEC review, the NRC intends to perform audit evaluations 'of feedwater line breaks, steam line breaks, and large- and small-break LOCAs (as part of the FSAR and TMI Action Plan Item II.K.3.30 and,II.K.3.31 reviews). In order to perform these audits, we reguire the following data, as outlined in the "PWR Information Request Package."
RESPONSE: The response will be provided on the CESSAR docket.
September 1981 15A.12-1 Amendment 6 07/30/81
PVNGS FSAR APPENDIX 15A (15.0)
One of the key parameters in LOCA analyses is peak clad temper-ature. For non-LOCA transients, minimum DNBR (departure from nucleate boiling ratio) is of primary importance. For those transients analyzed in Section 15 of the FSAR, provide graphical output of the DNBR as a function of time.
RESPONSE: The response will be provided on the CESSAR docket.
September 1981 15A.ll-l Amendment 6 07-30-81
PVNGS FEAR APPENDIX 15A fl""'": (" Q (15.0)
Plant operators are instructed to trip the reactor coolant pumps (RCPs) during ECCS actuation. For a steam line break, stripping of the RCPs at varying times into the transient has not been addressed. Demonstrate, by analysis or otherwise, that the consequences of tripping the RCPs during a steam line break transient are bounded by the analyses already performed.
RESPONSE: The response will be provided on the CESSAR docket.
September 1981 15A.10-1 Amendment 6 07-30-81
PVNGS FSAR APPENDIX 15A (15.0)
For each accident, discuss non-safety grade equipment which was
'I assumed to operate and could result in the transient becoming more severe or verify that no non-safety grade equipment operat-ing would produce a more severe transient. For example, the pressurizer heaters being energized for a transient resulting in high RCS pressures could tend to worsen the effects of the transient. Likewise, pressurizer spray could be deterimental for a transient resulting in low RCS pressure.
RESPONSE: The response will be provided on the CESSAR docket.
September 1981 07-30-81'mendment 15A.9-1 6
,~
PVNGS FSAR APPENDIX 15A OUESTION 15A.S (NRC Question 440.36) (15.0)
Verify that for each transient analyzed in Chapter 15, if operator action is not discussed then no operator action is required. In particular, consider events in which the ECCS is actuated or RCP trip would be required based on present procedures.
RESPONSE: The response will be provided on the CESSAR docket.
September 1981 15A.8-1 Amendment 6 0/-30-81.
PVNGS FSAR APPENDIX 15A OUESTION 15A.7 (NRC Question 440.35) ~ (15.0)
The method that you have used for calculating the amount of failed fuel'fter an accident has not been approved. It is our position that fuel failures be recalculated using the criteria that any fuel rod which has a CE-1 DNBR less than the minimum DNBR value determined in Section 4.4 fails. Radiological con-sequences should be calculated accordingly.
RESPONSE: The response will be provided on the CESSAR docket.
September 1981 15A.7-1 Amendment 6 07.-30-81
PVNGS FSAR APPENDIX 15A Our.STION 15A.6 (NRC Question 440.34) (15.0)
Confirm that during the preoperational or startup test phase you intend to verify the valve discharge rates and response times (such as opening and closing times for main feedwater, auxiliary feedwater, turbine and main steam isolation valves, and steam generator and pressurizer relief and safety valves) to show that they have been conservatively modeled in the Chapter 15.0 analyses.
RESPONSE
~~peak 3~ v'eri4q v-espouse 4'>1S Q glqoLo ~k +4y 4<ve. be.ev Co~ser vakiYel
~goIeIeg iw Cgqrker l S.O a~lqses during py-eo]era b~a es as eleicriM iw CC~SAR. C4(rfer l+ S~ valves ikey'n g4, CEKA1Z, S cop~ mZ PIIA6$ FSQ2
~
L C4l M l4 & Valves ouhrde CEO'cythe.,
September 1981 15A.6-1 .
Amendment 6 07/30/81
PVNGS FSAR APPENDIX 15A QUESTION 15A.5 (NRC Question 440.33) ~ (15.0)
For all analyses of transients with concurrent single failures, provide a reference to the sensitivity'study which shows that the failure selected is the worst case single failure.
RESPONSE: The response will be provided on the CESSAR docket.
September 1981 15A.5-1 Amendment 6 07-30-81
PVNGS F SAR APPENDIX 15A QUESTION 15A.4 (NRC Question 440.32) ~ ~
(15.0)
Expand Table 15.0-6, the list of single failure considered in transient and accident analyses, to include the following:
one primary safety valve stuck closed
~
C
\~ one secondary safety valve fail to open or fail to close 3 ~ loss of offsite power
~
M~
I failure of one diesel to operate (for the events with loss of offsite power being treated as a consequential result of the event).
failure to achieve fast transfer RESPONSE: The response will be provided on the CESSAR docket.
September 1981 15A.4-1 'Amendment, 6 07-30-81
~ .
\
~ It-. (6-3)
Provide a commitment that Palo Verde will perform tests of ECCS as installed to confirm that the actual ECCS flow rates are greater than the values assumed in the LOCA analyses.
RESPONSE
<<d,u > CasW Cpl l$
pgIJGs oil( )er4>~ <esses og Ecc5nfo
~V ~
Cog(iy-n, g~
+~
grmikev ~ ack~( ECCS -Ao~ <<~~~
Qe Yalues <$ $~~'~ l~
LocA a~~l~ve~. d~~~g~ ~ I'<<g~~ ""~
September 1981 6A.48-1 Amendment 6 07<<30-81
PVNGS FSAR QUESTION 6A.47 (NRC Question 440.30) ~ (6.3)
Provide a commitment that Palo Verde will perform preoperational and startup tests to meet the requirements of Regulatory Guide 1.68 and 1.79.
6 \
RESPONSE
Q'L IPp h BUNGS ~i]( Per4rin preqevnfioiia( a~8'Arf'uf keg'o km~ Qi Mguirevneu& of Peju(w+ 6uih's I.68 <<>>g (.7) as ouk(i v~8'i'n CESAR. (ha(iter I+
& /esses iz CESZjgzcy.e. a J PYRIC)S FZhR Ctiq>fel-l+ Ar+esM ou6ida eg CeW+g. 5cpe.
,September T
1981
~. ~
k 07-30-81 'I 6A.47-1 Amendment 6 I6-
Describe the instrumentation available for monitoring ECCS performance during post-LOCA operation (injection mode and recirculation mode) .
RESPONSE: The response will be provided on the CESSAR docket.
4
- ' r ~ '
'\*
September 198l. '----":-'7~3P ..i Amendment 6 8l ~ <.. ~
'~ 6" ( Q""""" (6.3)
Describe the means provided for ECCS pump protection including instrumentation and alarms available to indicate degradation of ECCS pump performance. Our position is that suitable means should be provided to alert the operator to possible degradation of ECCS pump performance. All instrumentation associated with monitoring the ECCS pump performance should be operable without offsite power, and should be able to detect conditions of low discharge flow.
RESPONSE: The response will be provided on the CESSAR docket.
September 1981 .".:,.'.,;.".;::,. ',;, 6A.45-1 "' .;..-: . Amendment 6
. 07-30-81
Recent plant experience has identified a potential problem regarding the operability of the pumps used for long-teM cooling (normal and post-LOCA) for the time period required to fulfillthat function. Provide the pump design lifetime (including operational testing) and compare to the continuous pump operational time required during the short- and long-term of a LOCA. Submit information in the form of tests or operating experience to verify that these pumps will satisfy long-term requirements.
RESPONSE: The response will be provided on the CESSAR docket.
~ n Septe~er 1981,:::. ":..:.;-".'. 6A.44-1:: - :-- .A ena ent 6 O7-30-81
~. 'IpI~g, P p 1 r
."UESTIGN 6A.43 (NRC Question 440.26) ~ (6.3)
Describe-. the instrumentation for level indication in the con-tainment'mergency sump., Also, provide detailed design drawings 8'f the containment emergency sump including the design provisions which preclude the formation of air entraining vortices during recirculation cooling. Confirm that the containment emergency sump design meets the requirements of'Regulatory Guide 1.82.
RESPONSE: Containment level instrumentation is provided to ensure there is sufficient net positive suc'tion head (NPSH) for the safety injection pumps and to verify that essential equipment is not flooged. Th range provided is from plus
. 6 in (gl. g0.5 es abave the sumps>to p ca)) us 6 inches above the maximum
/,5 fcePj floo eve3. A total range of eleven feet is provided in the control room. This safety-grade instrumentation is redundant, physically separated, environmentally qualified to post-LOCA environment, seismically qualified to function during and following a safe shutdown earthquake (SSE) and powered from redundant Class IE sources. The containment emergency sumps and screens are designed in full compliance with NRC Regulatory Guide 1.82, Revision 0. The ~sumps hydraulic performance was tested on a one-to-one>model scale in .
a hydraulic laboratory. As a result of the tests, a ArecCnd special vortex breaking cage will be installed<~the safety injection sump suction line inside each sump.
The tests have shown that the hydraulic performance of the sump is satisfactory with the vortex breaking cage installed. Further information on the model study is contained in the transcript to the Containment Systems Independent Design Review submitted under PVNGS transmittal letter ANPP-18147, dated June 4, 1981.
(<~7 i 4'ed)
~j5 y gnyi5 AtG.YB &6, VnihiNulA level $ >v. MPA ~ L~i".~~~~<
September 1981 6A.43-1 08-06-81 ,Amendment 6
In the event of early manual reset of the safety injection (6.3) actuation signal (SIAS) followed by a loss of offsite power during the injection phase, operator action may be required to reposition ECCS valves and restart some pumps. The staff requires that operating procedures specify SIAS manual reset not be permitted for a minimum of 10 minutes after a LOCA.
Provide the administrative procedures to ensure correct load application to the diesel generators in the event of loss of offsite power following an SIAS reset. h
RESPONSE
'r, ~h,e'y ip The. SIAM'an on( be rese~ when We inikia>iag grume(ers have
~e have. been resk ~
clea.l ec{. ~p gf S (vere. reset,,+km
-%%d nome(
~ufZ no( be in +4. lnJeckbn kNcde bu+ H1e
~na. %Au. sak$ in(ed ~~ cyst~
cav)dikes mould Ak <nJec&0H putts auld( continue. A operute un&i( individual <<u+
Q +e OPcrakoe.
pli'ggS'ocedures rovide'suAicieut iA4rrn+on +o
~+
minimum +he. o~ ~
un'((
f can eke-+he.]ro~'<<hl)w 'to rester (a f+o a. ~e, condition. P KIAZ ~'Il no/be reset ccn/ess ~ ~y Ias defErlo>nzcE +4+ ~JIAOns ~arran+
~;s ac$ y~. J.w ae(di4>n, +ruin~ proces'ur7es Mi'iiSP'<<$
~Q S(AZ nxtnuaf YeseP 'no& be o.permi8eZ ~<c~
op io ninukes uMr ~
~ill 6 p>>'cI~+ cole'r9
~krona(g(,
+e d(ese(
procedures general~.
ind<<i~) ~ri>>g ~
(lese proceduhU ~i(l ebs<<~ ~~ +
+Le. diesels av~ corred (y (oacLrd',
S(AS VM<~i
~A)e " ' 9+ n
'g (ass ag peg)er Q/(@urbe cc.
) 'vt = ~ ' ~ ~
1981'-'""'.'-"
n September . 6A.42-1 -' - " 'Amendment 6 07-30-83.
(
Assume a maximum passive failure flow rate of 50 GPM in each ECCS pump room and discuss the ~effects of the passiVe ailure to each ECCS pump operation, and demonstrate that adequate protection is provided for ECCS pumps from possible flooding.
RESPONSE
The SIS pumps are located at elevation 40 feet of the Auxiliary Building. Each pump is housed in a separate Seismic Category I reinforced concrete compartment.
RQ OP The leakage within each compartment is routed to>two separate train-related sumps. Each sump has its own pump. The embedded drain piping from each pump room to the respective sump is built to ASME Section III, Class 3 requirements. Each sump pump is cap&le oJ pumping 50 gal/min. Thevs&cu> g~e Cite. HO 4d~hfu,)
egeCk to EPICS Pu
~ ogrnk>n.
':;s..4MWiZ>WS'r sr's ~: .
s September 1981 Amendment 6 07-30-81
I! '*'
.23 llew!
~
discussion on specific methods of detecting, alarming and isolating passive ECCS failures during long-term cooling to include valve leakage. Show that there is sufficient time for the operator to take corrective action and maintain an (6.3) acceptable water inventory for recirculation. [1] Justify the basis for the assumed leak rates. [2] Describe how'the con-taminated water would be handled if one ECCS train must continue to operate with a leak.
RESPONSE
- 1. The response wi11 be provided on the CESSAR docket Addible(
difGcQ<On tS PeYi'dc4 iH +e. ref fenfe. Q hlkC QukS+0< 440s lb e
- 2. The leakage from the valves within the auxiliary building 'collected in radwaste sumps at the lowest building elevation of 40 feet. From this point, the wast~e ~
radwaste system (LRS) for processing.
can be pumped to the liquid YiZ(YC. (8e+e'5 Q>ggtd(e<ecL i~sgrnilicdnt odhev pzsEive. Scabs Aitsire<<Je.~y;giedL in ec~p<<ecl. ~ +tee ~ ~
+he. ~ezpsse. ~ NRC""gues4yp ~O. ((d.
September 1981,:::-" .."='...'.'."" 6A.40-1 !'mendment 6 07-30-.81 I<<
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TtIe ve'sgttse.~ill he. PntI'~ a~ 4u. CeZSWdon4+
F IIIJCs5'dan/gtt ~tt6 + 6~~8% Jltkf-7cca
. Ireguiretttett+s tde~A'fice( in .~. C6 $$ ~ I-eijrn~
I M addle)bn> +r i 1'em 6'., Vdr HeXing Feudenci'es
~rfAin +a +a~kgye Pic(uM bj a. s'cieWii Cage iuside '8ekt,'tk> similar in J.eszyn . k..ge InsQllecf in 4e. cog4at~~e~~an egregg sump.
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O. PVNGS FSAR amount of water above the suction pipes may also be unusable due to NPSH considerations and vortexing tendencies with the tank.
Preliminary indications are that approximately an additional 100,000 gallons of RWST capacity were needed to account for these considerations. It is our understanding that the design parameters for instrument error, transfer allowance and single failure have changed since the original sizing of the tank.
In light of the above information, discuss the adequacy of your Refueling Water Storage Tank. Provide a discussion of the necessary water volumes to accommodate each of the five considerations indicated above. Justify your choice of volumes necessary to account for each consideration. Provide drawings of your RWST, showing placement and elevation of tank suction lines, and level sensors. Also, provide operator switchover procedures for aligning to the recirculation mode, with estimates o'f the time required for each action.
/N~~T RESPONSE: 4(,
~l Amendment 6, '-'-'-'- '- .'. 6A.39-2 07-30-81 September 1981
Q '*' " ( Q"""""
PVNGS FSAR Recently, another plant has indicated that a design error existed in the sizing of their RWST. This error was discovered during a design review of the net positive suction head require-ments for the containment spray and residual heat removal pumps.
(6.3)
The review showed that there did not appear to be sufficient water in the RWST to complete the transfer of pump suctions from the tank to the containment sump, before the tank was drained and ECCS pump damage occurred.
It was reported that in addition to the water volume required for injection following a LOCA, an additional volume of water is required in the RWST to account for:
J ~ Instrument error in RWST level measurements.
C ~ Workin allowance to assure that normal tank level is sufficiently above the minimum allowable level to assure satisfaction of technical specifications.
0 ~ Transfer allowance so that sufficient water volume is available to supply safety pumps during the time needed to complete the transfer process from injection to recirculation. ~
- 4. Sin le failure of the ECCS system which would result in larger volumes of water being needed for the transfer process. In this situation, 'the worst single failure appears to be failure of a single ECCS train to realign to the containment sump upon low RWST signal. This result in the continuation of large RWST outflows and reduces the time available for the manual recirculation switchover, before the tank is drawn dry and the operating ECCS pumps are damaged.
- 5. Unusable volume in the tank is present because once the tank suction pipes are reached, the pumps lose suction and any remaining water is unusable. Additionally, some September 1981.:-"'-":; ";:.';;"'." 6A.39-1 Amendment 6 pg 30-81.,
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I I. The reactor operator will be performing those 'actions that < !
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' are required to be performed on the control panels. His actions will be guided by emergency procedures that will -.
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indicate what actions are required to be accomplished to ~ 4 I I I
correct or restore those parameters to minimize the con-
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h sequences 'and severity of the accident. ' ~ ~ C
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~, After preparation of the emergency procedures, PVNGS will perform a walk through of the procedure on its simulator and verify the operator has enough time to complete all ~ T required actions. *.'
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o PVNGS FSAR IL ' ( Q""""" (6.3)
Provide a time reference for each action in the sequence of action included in the changeover from injection to recircula-tion. Indicate the time required to complete each action and what other duties the operator would be responsible for at this point in the accident. How much time does the operator have to assure that the system is realigned to the recirculation mode before RWST water is exhausted if the RWSP isolation valves are not closed? Consider the required pump NPSH in your response.
If the operator fails to close the RWST isolation valves, demonstrate that the HPSI will continue to adequately cool the core during the recirculation mode.
RESPONSE
The response will be provided on the CESSAR docket.
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September 1981 6A.38-1 Amendment 6 07-30-81
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Provide in the Technical Specifications, (1) the range of nitrogen cover gas pressure for the SIT, and (2) the ECCS pump discharge pressures.
RESPONSE: The response will be provided on the CESSAR docket.
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C September 1981 ':, . -' ' 6A.37-1 Amendment 6 07-30-81
Provide the basis for ECCS lag times. Are these times calculated or verified by test. If calculated, are they verified during preoperational tests, and periodically reverified?
RESPONSE
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II September 1981 " -" -". s "
"" 6A.36-1 Amendment 6
7-30-81
Q A reported event has raised a question related to the conservatism of NPSH calculations with respect to whether the absolute minimum available NPSH has been considered. In the past, the required NPSH has been taken by the staff as a fixed number supplied through the applicant by either the architect engineer or the pump manufacturer. Since a number of methods exist and the method used can affect the suitability or unsuitability of a particular pump, it is requested that the basis on which the required NPSH was determined be branded (i.e., test, Hydraulic Institute Standards) for all the ECCS pumps and the estimated NPSH variability between similar pumps
'ncluding the testing inaccuracies be provided.
RESPONSE: The response will be provided on the CESSAR docket.
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I September 1981 ',";,' 6A.35-1 I~
I Amendment 6
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07-30-81
PVNGS FSAR APPENDIX 15A QUESTION 15A.13 (NRC Question 440.41) ~ (15.0)
The current CESEC model does not properly account for steam formation in the reactor vessel. Therefore, for all events in which (a) the pressurizer is calculated to drain into the hot leg, or (b) the system pressure drops to the saturation pressure of the hottest fluid in the system during normal operation, we require the applicant to reanalyze these events with an accept-able model or otherwise justify the acceptability of Palo Verde Chapter 15 analyses conclusions performed with CESEC.
RESPONSE: The response will be provided on the CESSAR docket.
September 1981 15A.13-1 Amendment 6 07/30/81
PVNGS FSAR APPENDIX 15A Figure 15B-19 shows the primary system pressure exceeding.ll0%
of the design pressure. This figure also indicates a substantial pressure differential between the pressurizer and reactor vessel.
The standard-review plans typically limit the pressurization of the RCS to 110% of the design pressure. However, the ASME pres-sure vessel code permits exceeding the 110% limit to approxi-mately 120% for very low probability events. The NRC will accept the limiting pressurization transient (i.e., feedwater line break) as calculated for System 80 if we can be assured that the analysis performed is conservative and that a small break in the feedwater line is a very low probability event.
As such, we request the following information be provided:
(1) Verification of CESEC to predict pressurization transients.
This should include the developed pressure differential across the pressurizer surge line.
(2) Demonstrate that the probability of a small break in the feedwater system is not significantly more probable than the large break. Include the consideration of ancilliary line breaks.
(3) Section 15B.3 references a sensitivity study for RCS overpressurization transient to plant initial conditions.
Provide the results to this study in graphical form.
Specifically, include DNBR and pressure as a function of time:
(4) It is expected that increasing the break area for a feed-water line break would increase the degree of primary system pressurization. 'A larger break area should result in an earlier loss of heat sink and corresponding higher decay heat for system pressurization. Figure 15B-1 indicates that the limiting feedwater line break is not a doubleended
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<<'s ji September 1981 15A.14-1 Amendment 6 07/30/81
PVNGS FSAR APPENDIX 15A guillotine break (1.4 ft ), but a 0.2 ft break. Provide greater details as to I
why this occurs. Is this behavior considered realistic or a consequence of a modeling assump-tion? Provide additional graphical explanations, including heat transfer coefficient, heat flow, secondary side inven-tory, all secondary side flow rates, and any additional data required to demonstrate the reasons for the 0.2 ft break being the limiting break size.
(5) Figure 15B-10 provides the relationship between the maximum RCS pressure to initial steam generator inventory. Provi'ce additional information which explains in detail functional behavior of this curve. Provide the RCS pressure curves for the cases of initial SG inventory of 95,000 and 170,000 ibm. Describe the SG heat transfer occurring throughout these events.
Page 15B-5 states: "...the initial RCS pressure can be adjusted to provide simultaneous reactor trip signals from high pressurizer pressure and low water level in the intact steam generator and hence the plateau of maximum RCS pressure." Provide greater details of the analyses and assumptions made in order to achieve coincident, trip signals from the pressurizer and SG.
(6) For Figure 15B-11 (and page 15B-6), how does raising the degree of feedwater subcooling increase the maximum RCS pressure? It would appear that raising the degree of subcooling would result in a larger heat sink, and, therefore, a lower peak pressure.
(7) What decay heat model does CESEC use? Does this model assume infinite irradiation?
(8) Provide details of the core and steam. generator heat transfer models used in CESEC.
I Amendment 6 15A.14-2 September 1981 0//30/81
PVNGS FSAR APPENDIX 15A (9) Utilizing a one-node representation of the steam generator secondary side, how is the low liquid level trip analyzed":
pressurization transients (resulting in the opening of a safety valve or PORV) with data from experiments and operating plant transients. Of interest is level and pressure as a function of time. Document the assumptions made in analyzing these tests.
(ll) Document the sensitivity of a feedwater line break with and without loss of offsite power.
RESPONSE: The response will be provided on the CESSAR docket.
S ep tember 19 8 1 15A.14-3 Amendment 6 o7/3o/sl
For the feedwater line break analysis, provide the pressurizer liquid and mixture level as a function of time.
Provide detailed plots for the following parameters during the initial 50 seconds of the transient:
Pressurizer Pressure Surge line flow
- 3. Pressurizer mixture level Pressurizer Safety Valve flow and quality RESPONSE: The response will be provided on the CESSAR docket.
September 1981 15A.15-1 Amendment 6 07-30-81
l PVNGS FSAR APPENDIX 158 UESTION 15A.16 (ivRC Question 440.44) (15.0)
Ve require additional information regarding, the steam generator behavior during a feedwater line break. Provide the steam gen-erator secondary side coolant inventory, mixture level, heat transfer coefficients, energy removed by each steam generator (Btu) and secondary side flow as a function of time.
It is our understanding that the limiting heat transfer modeling technique utilized in CESEC assumes an approximately constant heat transfer coefficient between the primary and secondary systems until all the liquid mass in the secondary system is depleted (i.e., bM = 0). It is not clear why the limiting modeling technique was not the case where the heat transfer was degraded as the secondary side inventory began uncovering the tubes. Please explain.
Discuss differences in the steam generator secondary heat transfer modeling between a feedwater line break and a steam line break.
RESPONSE: The response will be provided on the CESSAR docket.
September 1981 15A.16-1 Amendment 6 07-30-81
PVNGS FSAR APPENDIX 15A UESTION 15A.17 (NRC Question 440.45) (15.1.3)
The stuck-open atmospheric dump valve analysis assumed operator action to scram the core 1200 seconds into the transient.
Justify the time of manual action. Provide details of the plant symptoms which will alert the operator of the stuck-open dump valve. When will the plant automatically scram without operator invervention'? Discuss the failures assumed in the. analysis.
Question 440.41 addresses concerns with the capability of the CESEC code to properly account for primary system voiding.
Address the concerns of this position as they related to your analysis of the stuck-open atmospheric dump valve event.
Provide graphical output of the mass flow rate exiting the dump valve as a function ot time.,
When analyzing a stuck-open dump valve, operator action was required to isolate the feedwater from the affected steam generator. Justify the conservatism of time for operator action assumed in the analysis. Wnat signals do ihe operators receive signifying that the feedwater should. be isolated? When assuming tech-specs limits for the steam generator tube leakage, describe how CESEC accounts for the primary to secondary mass ~
depletion. In the analysis, the primary system was initialized to design operating conditions. Address the conservatism of this assumption when compared to off-nominal tech-specs limits and hot standby conditions.
RESPONSE: The response will be provided on the CESSAR docket.
September 1981 15A.17-1 Amendment 6 07/30/81 '
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PVNGS FSAR APPENDIX 15A QUESTION 15A.18 (NRC Question 440.46} ~ (15.0}
Accidents resulting in containment isolation also isolate the component cooling water to the reactor coolant pumps. This can potentially lead to RCP seal damage which may result in a LOCA.
Address the time availble for the operators to restore the coolant to the seals. Has consideration been given to not isolating component cooling water to th'e RCP seals on contain-ment isolation? If pump seal integrity cannot be maintained, evaluate the consequential failure of the pump seals for the limiting accident.
RESPONSE: The response will be provided on the CESSAR docket.
September 1981 15A.18-1 Amendment 6 07-30-81
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PVNGS FSAR APPENDIX 15A QUESTION 15A.19 (NRC Question 440.47) ~ (15.1.4)
Section 15.1.4.2 addresses small steam line breaks outside-con-tainment (SSLBOC). The following questions relate to this section:
(1) Justify why a SSLBOC is limited to 11.5% of full power turbine flow.
(2) Update Table 15.1.4.2-1 to include Safety Injection Tank (SIT) initiation time. Also, provide SIT and HPI flow as function of time ~
(3) During a small steam line break, the reactor core initially responds to a load demand. What break size results in the highest power excursion'? For the limiting break size, provide graphical output, of the system pressure, core power, and DNBR as a function of time.
(4) Explain why the liquid mass within the broken steam gen-erator increases after 1080 seconds. Isn't the steam gen-erator isolated? If not, why not?
(5) Why was the open dump valve accident (Section 15.1.3) analyzed at full power and the small steam line break (Section 15.1.4) analyzed at zero power? Assuming a tech-spec steam generator tube leakage of 1 gpm for both analyses, why wasn't the resulting dosage the same?
(6) What was the single failure assumed for the small steam line break? Justify the single failure selection as resulting in the limiting conditions.
(7) Provide graphical output of the ECC flows as a function of time and indicate when boron began to penetrate the primary system. How is the time to boron injection derived?
(8) Address the consequence of loss of AC power during the transients analyzed.
September 1981 15A.19-1 Amendment 6 07/30/81
'PVNGS FSAR ]I i4; APPENDIX 1SA 8 8 E'I
{9) Question 440.41 addresses concerns with the capability of ~W the CESEC code to properly account for primary system rV, voiding. Address the concerns of this question as they relate to your analysis of small steam line break outside of containment.
{10) Provide diagrams of the reheater offlines (include dimensions, loss coefficients, interconnections between 'p the steam lines). This data should be sufficiently detailed to enable the NRC to conduct an audit of a steam line break coincident, with a failure of an MSIV to close. Provide results for this accident (i.e., system pressure, pressur-izer level, DNBR ratios, ECCS flows, steam generator flows, etc.) assuming with and without loss-of-offsite power. Address the consequence of losing offsite power during the steam line t
break.
{11) Analysis of an inadvertent opening of a turbine bypass valve has not been provided. For this accident, will the DNBR fall below 1.19 as it did for Waterford'? If not, discuss the differences between the plants which cause the DNBR limit to be exceeded for one plant and not. the other. If the DNBR limit is exceeded, provide a detailed
, analysis for this event.
Provide a list of all accidents (excluding primary system LOCAs) which result. in a DNBR less than 1.19.
{12) Compare the steam flow model utilized in CESEC with the Moody slip flow model.
RESPONSE: The response will be provided on the CESSAR docket.
Amendment 6 15A.19-2 September 1981 07-30-81
PVNGS FSAR APPENDIX 15A I! -* ~ . ( ~ tt-. (15.0)
Provide a list of transients which result in opening of the pressurizer safety valves.
RESPONSE: The response will be provided on the CESSAR docket.
September 1981 15A.20-1 Amendment 6 07<<30-81 w~ g
PVNGS FSAR APPENDIX 15A QUESTION 15A.21 (NRC Question 440.49) u (15.0)
The staff has been informed that the CESEC-III computer program is best suited to analyze transients which void the upper head of the reactor vessel. As such, we request that the following information be provided:
(1) Documentation of the CESEC-III code. As part of the documentation, address the differences between the different versions of CESEC (I, II, and III).
(2)'rovide comparative analyses with the different versions of the CESEC programs (used for licensing) to demonstrate the adequacy of previous analyses.
(3) Provide verification of CESEC-III against plant and experimental data for pressurization and depressurization transients (such as the ANO-2 experiments and the St. Lucie I cooldown experience).
(4) For those transients which result in primary system voiding, provide graphical output of the upper head mixture level as a function of time. Discuss operator actions/guidelines for detecting and mitigating primary system void formation. >
(5) Show, by analysis or otherwise, that the allowable cooling rate (for cold shutdown conditions) will not result in primary system voiding.
RESPONSE: The response will be provided on the CESSAR docket.
September 1 9 81 15A.21-1 Amendment 6 07/30/81
PVNGS FSAR APPENDIX 15A Q '(15.0)
Do all CE steam generator designs incorporate a flow restrictor in the steam generator outlet nozzles?
RESPONSE: The response will be provided on the CESSAR docket.
September 1981 15A.22-1 Amendment 6 l6 07/30/81
PVNGS FSAR APPENDIX 15A (15.0)
Section 15C.3.1.3.3 is confusing. Provide greater detail of he reactor vessel mixing model. How do the various versions of CESEC evaluate asymmetric temperatures between the loops during a FWLB and a SLB (assuming with and without loss of offsite power)? Provide experimental verification for these models.
RESPONSE: The response will be provided on the CESSAR docket.
September 1981 15A.23-1 Amendment 6 07/30/81
PVNGS FSAR APPENDIX 15A QUESTION 15A.24 (NRC Question 440.52) (15.0)
Section 1SC.3.3 implies that during a SLB, concurrent with loss-of-offsite power, the reactor trips on a low DNBR 4ignal.
It is our understanding that CESEC does not calculate DNBR.
How is the time of reactor trip calculated?
RESPONSE: The response will be provided on the CESSAR docket.
September 1981 15A.24-1 Amendment 6 07/30/81
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PVNGS FSAR APPENDIX 15A Q""""" (15. 0)
The inadvertent opening of an atmospheric dump valve event is considered as a moderate frequent event per SRP 15.1.1. - Confirm that the analysis performed for this event in Section 15.1.3.2 is the limiting case identified by a qualitative comparison from the events in the same category group specified in SRP 15.1.1 (e.g., decrease in feedwater temperature, increase in feedwater flow, increase in steam flow, and inadvertent opening of a steam generator relief or safety valve). The qualitative analyses for each of the events in this group should be presented in the FSAR for staff review. Also, the results of analyses should be presented in the FSAR for each event with their worst single failure combination and the limiting case is identified.
RESPONSE: The response will be provided on the CESSAR docket.
.September 1981 1SA.25-1 Amendment 6 Qj/30/81
PVNGS FSAR APPENDIX 158 OUESTION 15A.26 (NRC Question 440.54} (15.0}
The depressurization transients analyzed for System 80 were conducted utilizing the CESEC-II computer program. This program does not account for steam formation in the upper head of the reactor vessel nor for steam formation in the primary system after the pressurizer empties. Neglecting these effects can esult in the improper evaluation of the system pressure and hydraulic behavior. The importance of this phenomenon was demonstrated by the St. Lucie I natural circulation cooldown event of June 11, 1980.
The modeling deficiency in CESEC-II described above has the potential for providing unacceptable results for the depressur-izing transients analyzed in the FSAR. As such, for all trans-ients which empty the pressurizer or may result in saturated conditions elsewhere in the primary system, the CESEC-II computer program must be verified to demonstrate it can correctly cal-culate svstem thermal-hvdraulic responses. The staff requires the applicant to demonstrate the acceptability of the CESEC-II program to properly account for the thermal-hydraulic phenomena in question, and to demonstrate compliance with NRC regulations.
In addition, we require a description of the SESEC code's ability to calculate the asymmetric cooldown between the intact and broken loops. Overlay plots of the hot leg and cold leg tempera-tures in the intact and broken loops should be provided.
RESPONSE: The response will be provided on the CESSAR docket.
September 1981 15A.26-1 Amendment 6 07/30/81
PVNGS FSAR APPENDIX 15A For small-break LOCAs, containment isolation may occur. It is our understanding that component cooling water to the RCP seal="
will be isolated upon containment isolation. Demonstrate that the RCP seals will remain intact and maintain the pressure boundary for the duration of the accident. Address expected RCP operation. If seal integrity cannot be maintained, seal failure must be assumed. Discuss the maximum seal leakage "ates based on operating experience. If the consequences of seal failure are assumed to be covered by the analyzed break spectrum, justify the differences in the break locations from the locations analyzed.
RESPONSE: The response will be provided on the CESSAR docket.
September 1981 15A.27-1 Amendment 6 07/30/81
The LOCA break spectrum analyses presented are stipulated to be applicable to any System 80 plant that conforms to the interface requirements specified within Section 6.3.3. The submittal for the LOCA analyses does not address the effects of steam generator tube plugging. The effect, of a decrease in steam generator tube flow area is an increase in the peak cladding temperature (when the peak occurs during the reflood portion of the transient). If the analyses provided are considered to support generators with plugged tubes, describe the intent of the plugging the analyses'support and the method used to account for the plugging. If steam generator tube plugging was not considered, the applicant will be required to perform additional ECCS analyses prior to operation with plugged generator tubes. In either case, the applicant is required to include an interface requirement on the validity of the LOCA'nalysis (acceptance criteria of 10 CFR 50.46) and the Technical Specification limit for the number (or percentage) of allowable plugged steam generator tubes.
RESPONSE: The response will be provided on the CESSAR docket.
September 1981 6A.49-1 Amendment 6 07-30-81
PVNGS FSAR APPENDIX 15A OUESTION 15A.28 (NRC Question 440.57) (15.6)
In light of recent operating experiences (the St. Lucie Unit 1 natural circulation cooldown event of June 11, 1980, and re-analyses of SAR Chapter 15 design bases events by St. Lucie in February 1981) a potential deficiency has been identified with the CESEC computer program and NSSS model. As the press-urizer cools down and the system pressure decreases, steam can form in the reactor vessel upper head due to flashing of the hot coolant in this stagnant region. The steam bubble in the reactor vessel upper head displaces coolant from the reactor vessel into the pressurizer and the steam in the vessel head will determine the system pressure. The CESEC model used for the steam generator tube rupture event does not account for this occurrence. Further, CESEC analyses which predict that the pressurizer will empty, or that the reactor coolant system saturates, do not appear to correctly calculate the system thermal-hydraulic response and are not justified for use.
Tnese events are to be re-analyzed witn a suitable model or additional justification is to be provided for the CESEC analyses to demonstrate that the computer program conservatively accounts for the formation of steam in the reactor coolant system.
RESPONSE: The response will be provided on the CESSAR docket.
September 1981 15A.28-1 Amendment 6
'07/30/81
PVNGS FSAR APPENDIX 15A (15.6.2)
The analysis for a steam generator tube rupture does not address tube leakage in the unaffected steam generator. Provide an interface requirement for the allowable steam generator tube leakage and reference the Technical Specification limit. Conf-'rm the analyses were performed using this allowable limit or provide justification why this leakage term can be excluded from the analyses.
RESPONSE: The response will be provided on the CESSAR docket.
September 1981 15A.29-1 Amendment 6 07/30/81
PVNGS FSAR APPENDIX 1SA Q" (15.6.2)
The analysis for a steam generator tube rupture is for a double-ended rupture. Provide the analyses used to determine that this is the limiting ease. If a partial area break is considered, such that the steam generator relief valves open at a longer time into the transient is more primary coolant leaked to the secondary and out the SRVs, resulting in an increased dose rate.
RESPONSE: The response will be provided on the CESSAR docket.
September 1981 15A.30-1. Amendment 6 07/30/81 I ~
PVNGS FSAR QUESTION 16A.31 (NRC Question 440.60) ~ APPENDIX 15A (15.6.2)
SRP 15.6.3 acceptance criteria requires that this event be analyzed with a concurrent loss of offsite power. Provide an analysis for the limiting case which includes a concurrent loss of offsite will be power.'ESPONSE:
The response provided on the CESSAR docket.
September 1981 15A.31-1 Amendment 6 07/30/81
PVNGS FSAR APPENDIX 15A (15.6.2j For the SGTR event, what prevents steam from the affected steam generator b'eing used to drive the steam-driven auxiliary feed-water pump -and exhausted to the environment? If operator action is required, confirm that no credit for operator action was given for 30 minutes, consider with your assumption for isolation of the affected steam generator. If credit was given for operator action in less than 30 minutes, provide justifica-
, tion why this credit can be given, or reanalyze the event. assum-ing steam from the faulted steam generator is used to drive the steam-driven AFW pump and is exhausted to the environment.
RESPONSE: The response will be provided on the CESSAR docket.
September 1981 15A.32-1 'mendment 6 07/30/81
PVNGS FSAk
- APPENDIX 15A QUESTION 15A.33 (NRC Question 440.62) (15.6.3,4,5)
Provide a description of the CESEC model used to model the. CVCS from the reactor coolant system to the break point. Include a description of the environmental conditions at the'reak point (pressure, enthalpy, break flow model used).
RESPONSE: The response will be provided on the CESSAR docket.
15A.33-1 Amendment September 1981 6 07/30/81
PVNGS FSAR APPENDIX 15A QUESTION 15A.34 (NRC Question 440.63) ~ (15.6.3,4,5)
Discuss the single failure assumed for'hese analyses. What analyses/evaluations were performed to justify that the- single failures chosen were the most limiting'P RESPONSE: The response will be provided on the CESSAR docket.
September 1981 15A.34-1 Amendment 6
PVNGS FSAR APPENDIX 15A uESTION 16A.36 (hsC Question 440.64) (15.0j In this section, you have selected the turbine trip without a single failure as the limiting reactor coolant system pressure ~
- and the limiting radiological release event for the moderate frequent event category in the decreased heat removal by secondary system group. However, these limiting cases were not selected by a qualitative comparison of similar initiating events specified in SRP 15.2.1 through SRP 15.2.7 (e.g., loss of external load, turbine trip, loss of condenser vacuum, steam pressure regulator failure, loss of normal AC power and loss of normal feedwater flow). -Provide a qualitative analysis in the FSAR for each of the initiating events in the same group per the SRP, and identify the limiting cases for the group. Provide a detail quantitative analysis for each of the limiting cases including the limiting RCS pressure, limiting fuel performance, and the limiting radiological release.
RESPONSE: The response will be provided on the CESSAR docket.
15Ae35-1 Amendment 6 September 1981 07/30/81 4
PVNGS FSAR APPENDIX 15A.36 (NRC Question 440.65) ~ (15.2) 15'UESTION In this section, you have provided the loss of condenser vacuum with a fast transfer failure and technical specification steam generator tube leakage as the limiting RCS pressure and the the limiting fault limiting radiological release event for encases event category in the decreased heat removal by secondary system group. Although, these limiting may be the candidates for the limiting cases for the infrequent event category in the group, they were not selected by a qualitative comparison of similar initiating events plus a single failure specified in SRP 15.2.1 through 15.2.7. Provide a qualitative analysis in the FSAR for each of the initiating event plus a single failure in the same group per the SRP, and identify the limiting cases for the group. Provide a detailed quantitative analysis for each of the limiting cases including the limiting RCS pressure, limiting fuel performance, and the limiting radiological release. Confirm that the results of the analyses meet the acceptance criteria for these events per SRP 15.2.1.
RESPONSE: The response will be provided on the CESSAR docket.
September 1981 15A.36-1 Amendment 6 1
l6 - .
07y30/81 * .,IQ
PVNGS FSAR APPENDIX 15A Q" (15A)
Provide tabulations of the seguence of events, disposition of normally operating systems, utilization of safety systems, and a transient curve of primary system pressure for the total loss of primary coolant flow event. Also provide an analysis of the total loss of primary coolant flow with a single failure event.
Confirm that the results of these analyses meet the acceptance criteria for these events per SPP 15.3.1.
RESPONSE: The response will be provided on the CESSAR docket.
September 1981 15A.37-1 Amendment 6 07/30/81
PVNGS FSAR APPENDIX 15A OUESTION 15A.38 (NRC Question 440.67) (15.3)
In Section 15.3.5 you have provided the single reactor coolant pump shaft seizure with loss of offsite power following- turbine trip and with technical specification tube leakage as the limiting RCS pressure and radiological release event for the limited fault event category. This postulated event, is classified as an infrequent event'per SRP 15.3.3. Confirm that the results of the analysis meet the acceptance criteria for these events per SRP 15.3'.3, using the criteria stated in Question 440.35 to seamount calculate the amount of failed fuel in this event. State the of failed fuel in the results of the analysis. Radio-logical consequences should be calculated accordingly.
RESPONSE: The response will be provided on the CESSAR docket.
September 1981 15A.38-1 Amendment 6 l6 07/30/81
PVNGS FSAR APPENDIX 15A OUESTION 15A.39 (NRC Question 440.68) (15.0)
Provide results of an analysis of the reactor coolant pump shaft break as required by SRP 15.3.4 for staff review. The event should consider loss of offsite power following turbine trip and with technical specification steam generator tube leakage.
The criteria stated in Question 440.35 should be used for the calculation of the amount of failed fuel for this event. State the amount of failed fuel in the results of the analysis. Radio-logical consequences should be calculated accordingly. Confirm that the results of the analysis meet the acceptance criteria for these events per SRP 15.3.4 which classifies this event as an infrequent event.
RESPONSE: The response will be provided on the CESSAR docket.
September 1981 15A.39-1 Amendment 6 07/30/81
PVNGS FSAR APPENDIX 15A QUESTION 15A.40 (NRC Question 440.69) M (15.5)
In this section, you have provided the pressurizer level control system malfunction (PLCSM) with a fast transfer failure and the PLCSM with a loss of offsite power at turbine trip with techni-cal specification steam generator tube leakage as the limiting RCS pressure and radiological release event for the limiting fault event category in the increase in reactor coolant system
'nventory group. However these limiting cases were not selected by a qualitative comparison of similar initiating events plus single failure specified in SRP 15.5.1 (e.g., inadvertent operation of high pressure ECCS or a malfunction of the CVCS).
Provide a qualitative analysis in the FSAR for each of the initiating events (with and without a single active failure)
'n the same group per the SRP, and identify the limiting cases for the group. Provide a detailed quantitative analysis for ach of the limiting cases including the limiting RCS pressure, limiting fuel performance, and the limiting radiological release.
Confirm that the results o the analyses meet the acceptance criteria for these events per SRP 15.5.1.
RESPONSE: The response will be provided on the CESSAR docket.
'September 15A.40-1 1981 Amendment 6 07/30/81
PVNGS FSAR APPENDIX 15A UESTION 15A.41 (NRC Question 440.70) ~ (15.0)
Provide tabulations of the sequence of events, disposition of normally operating systems, utilization of safety systems, and all necessary transient curves for the startup of an inactive reactor coolant pump event. The comparison to peak RCS pressure acceptance criteria should be included in the analysis. Also pr'ovide the results of an analysis of this event with a single failure. Confirm that the results of these analyses meet the
-cceptance criteria for these events per SRP 15.4.4.
RESPONSE: The response will be provided on the CESSAR docket.
September 1981 15A.41-1 Amendment 6 07/30/81
~ .
PVNGS F SAR APPENDIX 15A OUESTION 15A.42 (NRC Question 440.71) (15.D)
You have provided, in Section 15D, the results of an inadvertent boron dilution event without a single failure under plant cold shutdown conditions. This information is not sufficient. You should provide results of analyses for all possible boron dilution events under various plant operational modes (e.g.,
refueling, startup, power operation, hot standby and cold shut-down). Also provide the results of analyses of these events with a single failure. Confirm that the results of these analyses meet the acceptance criteria for these events per SRP 15.5.1. In particular, the available times per operator action between time of alarm and time to loss of shutdown margin should be shown to meet the SRP guidelines. The results of the analyses should be presented in the FSAR including tabulations of sequence of events, disposition of normally operating systems, utilization of safety systems, and all necessary transient curves for the events.
In your analysis, indicate for all modes of operation what alarms would identify to the operators that a boron dilution event was occurring. Consider the failure of the first alarm.
Provide the time interval. from this alarm to when the core would go critical. If a second alarm is not provided, show that the consequences of the most limiting unmitigated boron dilution event meet the staff criteria and are acceptable.
RESPONSE: The response will be provided on the CESSAR docket.
September 1981 15A.42-1 Amendment 6 nv/sa/81
PVNGS FSAR APPENDIX 15A W- tt- (15.2)
As explained in Issue No. 1, NUREG-0138, credit is taken for closure of nonsafety-grade valves such 'as turbine stop valves, control valves, and bypass valves downstream of the MSIV to limit blowdown of a second steam generator in the event of a steam line break upstream of the MSIV. In the Palo Verde, Plant desi n there are various flow aths located between the MSIV and the turbine stoo valves Fi re 10.2-4 that serve various unidentified functions. To confirm satisfactory performance after a steam line break provide the following information, as applicable, related to these various flow paths that branch off between the MSIV's and the turbine stop valves:
(1) the function of the various flow paths and their maximum steam flow (2) the type of valves (3) the size of valves (4) the quality of the valves (5) design code of the valves (6) the closure time of the valves (7), the actuation mechanism of the valves (8) the closure signal including sensor (9) quality of power sources to valves and sensors (10) quality of air supply to air-operated valves (11) identify the valves that will remain open during main steam isolation In addition, provide justification or analysis that the failure of an MSIV and the additional blowdown paths result in a less severe accident than that analyzed in Chapter 15.
~g;@><~I >'NQ~on ye5p156 RESPONSE: The-~pere+c~> is contained in the amended pep~~
to question 10A.9 (NRC Question 430.45).
PvNGs pecks A~ ivrkvfac~ require~ ic(ceA SA CESS+ y~~e. 15A.43-1 id'e.
)e September 1981 Amendment 6 08-04-81
I ~ ~ ( - 1 ~ ~
PVNGS FSAR >A7W rZPENDIX 10A Q *1 ~ . 5) (10.3)
As explained in issue No. 1 of NUREG 0138, credit is taken for all valves downstream of the Main Steam Iso'ion ValVe (NSIV) to limit blowdown of a second steam generator in the event of a steam line break upstream of the M'IV. In order to confirm satis factory per formance following such a steam line break pro-vide a tabulation and descriptive text (as appropriate) in the.
FSAR of all flow paths that branch off the main steam lines between the MSIV's and the turbin= stop valves. For each flow path originating at the main steam lines, provide the following in form ation:
a) System identification b) Maximum steam flow in pounds per hour c) Typ'e of shut-off valve(s) d) Size of valve(s) e) Quality of the valve(s) f) Design code of the valve(s) g) Closure time of the valve(s) h) Actuation mechanism of the valve(s) (i.e , Solenoid operated, motor operated, air operated ii~phragm valve, etc.)
i) Motive or power source for the valve artuating mechanism In the event of the postulated acciden+, termination of steam flow from all systems identified abo', except those that can be used for mitigation of the acc'ent, is required to bring the reactor to a safe cold shut'~wn. For these systems describe what design features J.ave been incorporated to assure closure of the steam shut-of: ~alve(s). Describe what operator actions (if any) are requi =d.
If the systems that, can oe used for mitigation of the accident are not available or ~e.ision is made to use other means to Amendment 4 10A-6 Nay 1981
l PVNGS FSAR w ED3T C".: I".".:":
APPENDIX 10A shut down the reactor describe how these systems are secured to assure positive steam shut-off. Describe what operator actions (if any) are required.
If any of the requested information is presently included in the FSAR text, provide only the references where the informa-tion may be found.
RESPONSE: NUREG-0138 Page 1-9 states that the Probability of occurrence of the above scenario is quite low. Page 1-10 states that the scenario is not analyzed by the staff and need not be considered as a design basis accident. This scenario should therefore not be a design basis accident for Palo Verde Units 1, 2 6 3.
Refer to the following PSID's:
o 13-M-SGP-001 (figure 10.3-1) o 13-M-SGP-002 (figure 10.3-1)
~ 13-M-FTP-001 (figure 10.3-3) o 13-M-CDP-001 (figure 10.4-9) e 13-M-MTP-001 (figure 10.2-1)
~ 13-M-MTP-002 (figure 10.2-1) o 13-M-ASP-001 (figure 10.3-2) o 13-M-GSP-001 (figure 10.4-2)
Table 10A-3. lists the information requested. The table shows valve positions following MSIS isolation. For those valves which remain open, the total steam flow through these valves is 253,955 lb/h. Each auxiliary feedwater pump (AFM) has a capacity of 484,000 lb/h. Therefore, even for the extreme situation postulated, any auxiliary feedwater pump can prevent the second steam generator from boiling dry.
Q 0 . ( Q *I ~ .46) (10.4.1)
Provide a tabulation in your FSAR showing the physical character-istics and performance requirements of the maxn condensers. In August 1981 10A-7 Amendment 5 -,, l4 8-20-81
Table 10A-1 FLON PATHS ORIGINATING AT MAIN STEAM LINES (Sheet 1 of 2)
Nsu, Closer ~ clot lvs Syatan lt (dent lest lou
~ Scean rlo (Ls/nsl type ot Shut off valves Sire ot Vs Ice Oval f ty Ooslen of Valve Code Valve ot Tine of Valve (Second ~ )
Actuation NechahI ~ I ~
~rot Sou>ca Closure Si>chal (sensor)
Ouallty ot Ouellty of Posit)on Statue Pavo t Source ot Alc Supply of Valve Aft~ I NSIV Isolation Coe>>>ant 50 VOP) lt)$ Cate 0>> Hon-0 ANSI 15 Nshual N/A N/A N/A N/A . Open (or AS-Vool) (Cate) S)l.1 50 Votl 101$ Cate 0>> AU 51 1$ Nanual N/A N/A N/A N/A Auu stean supply lor AS-v011 (Cate) 0) I.I (I) n Asp Soll tl)-H-SCP-OO))
SC TOSS 105 Clobs )>> Noh-0 AN51 10 KshUOI 0/A N/A N/A N/A (or AS-Vol)) Icatol S)1,1 NT UV Iool O>1$ 510 Clobe P NISI 0,1 Nydraullc Trip of tur N5I$ Actuation Non-lt H/A Closed Ior w-loos)
'on S)1,1 bine speed conttol sye-IN>> (actuated on NSIS Sfunsl (Lov S/C
~ 5 ~ e ~ ur ~ I pecan>>tare)
ÃT W 1004 ~ >1SSI0 Clobo 1~ Non 0 NISI 0>1 Hydraulic Trfp of tur N515 Actuation Non-lt N/A Closed (or W 100)) S)1,1 bine speed $ )ens) ftov 6/C contcol sta- ~ reesuce) ten (actuated Hafn stean supply on NSIS In nein turb(ne parse>>tace) I I) N.NFP ool)
Ify UV 100) ~ >15510 Clobo 20>> Non-P ANSI ~ >1 Nydcsullc Tc(p of tur NSIS Actuation Non-lt H/A Closed (or W IOOI) S)1 ~ 1 bine speed S(Chal (Lov 5/C control ~ 're ~ co ~ ~ ucoI Ien (actuated estop on IISIS parse>>terai IIT W 1000 (ot UV-100))
~,)Sslo Clobo 10'on~ ANSI S)1,1
~ ,1 Hydraulic Trip to tut bine speed HSIS Actuation Slunal (Lov S/C Non-lt N/A Closed ronttol eye Pressure) t>>n (OC'IUO'ted on NSIS paraeetoce)
SC>>W-0)S SO.OOO I Clobe 1>> Non P ANSI 10 1 I Notor N/A Non-lt N/A S)l, I 50 W 0)i S0,000 'lobs Hon-0 ANSI 10 . Notor Hon IC>
I SOU, N/A Non lt N/A ~ Ised otf line betvesn NSIV' S)l. I 50 0)) $ 0>000 I Clobe Non~ .ANSI 10 Notoc phase> N/A Non lt H/A Open and turb(ne 50 UV W O)0 $ 0,004 Clobe Non-0 S)l ~ I ANSI Sf l,l lo, f Notoc 60 cycl ~
N/A Non-lt H/A Open valves (op>>ne on turbine trip)
(I) N SCP OOI)
PA U
M I
o
Table 10A-1 FLOW PATHS ORIGINATING AT MAIN STEAM LINES (Sheet 2 of 2) O CO Nas ~ Closure Not lvs Stean typo of Slat Shut off Cue llty Oee(OO Vlar of or Closure Ouollty of Oual(ty of ~ 0 ~ I(Ion St ~ tuo Syst>> P)ou o( of Code of Valve Ac'tuat(on Paver 6(usa I Poser of Alt of Valve Af let ldentlf Isation (LS/NS) valves Valve Valrt Valve (Seconds) Ntehahl ~ ln ShuICt (S>>sant) Coulee Supply NSIV I>>OI ~ 'tIOn Cosv>>tnt
~ 0 PV 1001 I ~ )io>>000 C)o'be 12>> 0 ANSI II)I, I 1$ Pa>>Is>>otic Non lt Ansi s)I I
~ Closed Nalh ~ 'ttan bleu dovh at ~ 'lsoe SO tV 1004 1,240,000 C)obo 12>> Non-0 ANSI O)l ~ I 1$ ~ nso>> I'I Ic Non lt NISI O)tel Closed ven'I tssttlc\or II)-N-SCP-00)I SO tV-1002 I ~ 240>>000 Clobe Non 0 ANSI 15 thole>>otic Non-It ANSI S)L ~ I Closed n) I ~ I SC tV 1001 I ~ 240,000 Clobe 12'2'2'1>>
Non 0 ANSI N)l. I 1$ Phelslo\Lc Insttu-sent Alr Solenoid parole olve to open Non lt ANSI S)1 I Closed lnoh IC 1)OV dc)
SC-tV 100) 1,240,000 Clobe 12'on Non 0 NISI S)l I 15 Patio>>otic scno parnleslro Non-It ANSI S)1 ~ I Closed Nein otean blou-dovn to eonutha ~
eluhsf Iotlc (I) N SCP 00)I SO PV 1004 1,1(0,000 Clobo ANSI the>>at(c lhon lt 1)OV ac) Non-LC ANSI O)I ~ I Clootd S)l ~ I SC-PV-IOO) 1,2(O,OOO Cl~ 12>> Non-0 ANSI S)1 I thtIO>>otic Non-lt ANSI S)LE I Clootd Cl 5
SC tr 1004 I ~ 140,000 Clobe NISI O)i >>1 15 tnouhst(C Non IC ANSI O)l.t Clostd yt NV 45 120,000 Clobo $>> Non 0 ANSI ~ >>S Nydcaulle pleat trip( )
Non-)E N/A Closed s)i>>1 NPN Punp NPN (or NY-41) l)OI 000 C lobe 5'>>
Non 0 ANSt ~ .) hydraulic turbine Lou(e (NSISI Non-lt N/A Closed Nein titan o)l ~ I Sp>>td ~ upply to tt NV-44 i)O.OOO Clobe Non 0 NISI P)i>>i
~ I) Nydraullc Control Systrn Non-Lt N/A Closed Wu pusp turbine lor NV-44) 120>>000'lobe 5>> ANS't ~ .) Nydcaulle Non )C N/A AC)osod P)I ~ I Nt UY )240 24)I $ 00 Clobe lo>> Non-0 Ahst 1$ heter Non-I t N/A CLossd n)I ~ I tltetr l Load etna(no Loo( Non-lt Closed Nein stean to hofatuco
~ uppl
)IIA Clobs 10>> Non 0 ANSI 15 Notor N/A Ny UY 242>>500
~ lt I
~ cal thon lt oh sa(h tutblho ~
II ~ .e prsaouco lt N/A aspatatot ethos'tor N1 UY )Ito 141, $ 00 Clobe lo>> Non 0 ANSI Notor deere
~ >>Ltch PSL 5l)I Non Closed
~ )I I
~ J phase Non<<lt Ny UV )20C 242>>)00 Clobo lo>> Non 0 ANSI 1$ Notor SO cyclo) N/A Closed n)LE I CS NV 005 50>>000 Cate ~>> Non 0 ANSI A)l ~ I Lo Notor tloctclc lhon IE, Control roon handavlteh Non-IC N/A Open (closed by plant op>>cater)
Nein ete>> supply
'to eland ~ oal o'Z
~ OOV
) phaste 40 cycle)
O g' ~. Nrn p>>p turb(ho trips oh hlth dteeharOO tressure duo to NPN lso)at(on rslres Volnt I hut on hain ~ tean Isolation elohal (NSIS) (Ior stean tehoretot pressure).
W
PVNGS FSAR APPENDIX 15A Several recent LERs indicate there has been a deficiency in the inadvertent boron dilution analysis at some plants. Provide an analysis of the dilution event when the RCS is drained to the hot leg.
RESPONSE: The response will be provided on the CESSAR docket.
15A.44-1 Amendment 6.
September 1981
S ~ - ~ ~ > M ~
PVNGS FSAR APPENDIX 15A UESTION 15A.45 {NRC Question 440.74) {15.D)
Recently, an operating PWR experienced a boron dilution incident due to inadvertent injection of NaOH into .the reactor cpolant system while the reactor was in a cold shutdown condition.
Discuss the potential for a boron dilution incident caused by dilution sources other than the CVCS.
RESPONSE: The response will be provided on the CESSAR docket.
September 1981 15A.45-.1 Amendment 6
PVNGS PSAR APPENDIX 15A OUESTION 15A.45 (NRC Question 440.74) (15.D)
Recently, an operating PWR experienced a boron dilution incident due to inadvertent injection of NaOH into the reactor coolant system while the reactor was in a cold shutdown condition.
Discuss the potential for a boron dilution incident caused by dilution sources other than the CVCS.
RESPONSE: The response will be provided on the CESSAR docket.
September 1981 15A.45-.1 Amendment 6 07/30/81
PVNGS FSAR APPENDIX 15A QUESTION 15A.46 (NRC Question 440.75) (15.6}
Discuss the transient resulting from a break of an ECCS injection line. In particular, describe the flow splitting which will occur in the event of a single failure and verify that the amount of flow actually reaching the core is consistent with the assumptions used in the analysis.
RESPONSE: The response will be provided on the CESSAR docket.
September 1981 15A.46-1 Amendment '6 07/30/81
PVNGS FSAR APPENDIX 15A QUESTION 15A.47 {NRC Question 440.76) ~ , (15.8)
The NRC is currently considering what actions may be necessary to reduce the probability and consequences of anticipated transients without Scram (ATWS). Until such time as the Commission determines what plant modifications are necessary, we have generally concluded that pressurized water plants can continue to operate because the risk from anticipated transient without scram events in a limited time period is acceptably small. However, in order to further reduce the risk from anticipated -tra'nsient without scram events during the interim period before completing the plant modifications determined by the Commission to be necessary, we have required that the following actions be taken:
Develop emergency procedures to train operators to recognize anticipated transient without scram events, including consideration of scram indicators, rod position indicators, flux monitors, pressurizer level and pressure indicators, prssurizer relief valve and safety valve indicators, and any other alarms annunciated in the control room with emphasis on alarms not processed through the electrical portion of the reactor scram system.
- 2. Train operators to take actions in the event of an anticipated transient without scram, including considera-tion of manually scramming the reactor by using the manual scram button, prompt actuation of the auxiliary feedwater system to assure delivery to the full capacity of this system, and initiation of turbine trip. The operator should also be trained to initiate boration by actuation of the high pressure safety injection system to bring the facility to a safe shutdown condition.
Describe how you will meet the above requirements, and provide a schedule for submit.ttal of the ATWS procedures for staff review.
INERT $ ~-
RESPONSE
September 1981 15A.47-1 Amendment 6 IG an/81
1 r Procedures will be developed to cover emergencies and off-normal events. -These procedures will provide suf-ficient guidance to ensure that correct action is tal cn by the operator. ATliS events will be covered in these procedures. PVNGS will provide training on hTHS events and emergency and off-normal procedures. Sufficient information will be provided so that the operator can if his actions are effective. Should the op- 'etermine erator's actions not be effective, the procedure will contain additional action that can be taken by the op-.
erator to ensure the parameter and/or condition is re-stored to acceptable values.
Procedures will, be available for NRC review at least 60 days prior 'to fuel load.
~ ~ ~
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List all. ECCS valve operators and controls that are located (6.3) below the maximum flood level following a postulated LOCA or main steam line break. If any are flooded, evaluate the potential consequences of this flooding both for short and long-term ECCS functions and containment isolation. List all control room instrumentation lost following these accidents.
RESPONSE: Air-operated drain valves SIB-UV-322 and 332 are used for relieving piping header pressure to the reactor drain tank after the reactor coolant system (RCS) check valve test> kd QC HS) aSed dS'iQ ggqygg 0+pfjch) ~
An air-operated containment isolation valve CHA-UV-560 is used to isolate the reactor drain tank discharge header.
A second isolation valve is located outside containment.
Pressure instruments SIA-PT-390 and SIB-PT-391 are used in conjunction with RCS check valve testing and can also be used for indication of check valve leakage.
No contr l room instrumentation is lost. +e ct$
qlo or s -te ood'f tern
'bov th s~ in w~ s
~i QeVe ~r~
gbgve kO haWJ i'A~X I -
~ On@
~ s~t-~~'ood7+
~<de i~J'e~4~
September 1981 6A.50-1 Amendment 6 07-30-81
lt" "
I'Because
( Q"""""
PVNGS F SAR of freezing weather conditions, blocking of the vent line on the refueling water tank (RWT) has occurred on at least one operating plant. Describe design bases and features that preclude this condition from occurring in the Palo Verde (6.3)
Plant.
RESPONSE: The refueling water tank (RWT) is provided with an eight-inch vent line that is connected to a .
common ten-inch header leading to the fuel building normal exhaust duct system. The water within the RWT will be kept above 60F at all times. The vent is located in the uppermost portion of the tank. The vent pipe is routed without piping pockets that could cause the accumulation of moisture. As the design winter ambient temperature at PVNGS is 25F for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, plugging of the RWT vent line is considered very improbable.
September 1981 6A.51-1 Amendment 6 07-30-81
PVNGS CESAR OUESTION 6A.52 (NRC Question 440.79) W- (6.3)
It is our position that the SIS hotleg injection valves should be locked closed with power removed during normal plant operation in order to prevent premature hotleg xngecta.on
~ ~ ~ 0 following a LOCA.
RESPONSE: Two valves in series are provided for each hotleg injection line. Each valve is powered from a separate power supply and is controlled by a keylocked switch in the control room. The design meets the single failure criterion to prevent premature hot leg injection.
September 1981 6A.52-1 Amendment 6 07-30-81
Your sump test program described in Section 6.2.2 is not in sufficient detail. The experimental program must demonstrate that sufficient margin in available NPSH over that required for each pump with all pumps at runout or maximum post-LOCA flow.
The test must demonstrate that the design precludes conditions adverse to safety system operation. Test parameters must include:
(1) minimum to maximum containment water level, (2) minimum to maximum safety system flow range in various combinations'this includes transients associated with startup, shutdown, or throttling of a train or pump), (3) random blockage of up to 50 percent of the screens and grids, (4) approach flow for each dominant direction and combinations thereof, and (5) simulation of break flow or drain flow impinging or originating within line of sight of the sump and its approaches.
If adverse conditions are encountered, the model configuration must be revised until an acceptable configuration is developed and demonstrated to perform over the full range of variables.
Since you choose to conduct a model test, provide details of the test program. Include information on the model size, scal-ing principles utilized, comparison of model parameters to expected post-LOCA conditions, and a discussion on how all possible flow conditions and screen blockages will be considered in the model tests. Whenever a reduced scale model is tested, all tendencies for vortex formation must be suppressed. Rota-tional flow patterns and surface dimples which might be accept-able in full scale tests, probably would not be accepted in a model program. Model testing must include some in-plant testing to demonstrate experimentally that NPSH margin exists for each peale.
RESPONSE: The test included a complete one-to-one~modeling of the system. This included various flow conditions and September 1981 '6A. 53-.1 Amendment 6 08-03-81
' vs A 4 IH ~ I ~
PVNGS FSAR, +sg gQ screen plugging by using a full scale model of the sump,
. screens, safety injection piping, instruments, and struc-tures in the sump vicinity. Further information on the model study is contained in the transcript to the Contain-ment Systems Independent Design Review submitted under PVNGS transmittal letter ANPP-18147, dated June 4, 1981.
lent"uA Tn cddik<ce, 4(e ApSB calculate'bn res<<lks are .
S<<>>o>>ar'3od bio~ + de>>ovs4ra4~ 4i.a+ suHi~ie +.
vnargie in availat>kl,.ipgH (s provi de'o( ovi'4h.af for e'Q.cg ttons f).
e Ma H~l 2.2. f~e+ 3 l.q 4'e.F g,g .gect-L'PS I l't $ eek- 3l 3 feW
~
t2.P gc~
CS 2. (p gee+ 53,4. 4'ee(- pe.eQ Lese Nag'the are ca(ca(a(eo( Qr Siinu((unco< <<)hick runout'lo<
those assumed in the NPSH calculations.~
Amendment 6 6A.53-2 September 1981 08>>03-81
'K
~
'.c OUESTION 6A.54 (NRC Question 440.81) (6.3)
During our reviews of license applications we have identified concerns related to the containment sump design and its effect on long-term cooling following a Loss of Coolant Accident {LOCA).
These concerns are related to (1) creation of debris which could potentially block the sump screens and flow passages in the ECCS and the core, {2) inadequate NPSH of the pumps taking suction from the containment sump, (3) air entrainment from streams of water or steam which can cause loss of adequate NPSH, (4) formation of vortices which can cause loss of adequate NPSH, air entrainment and suction of floating debris into the ECCS and (5) inadequate emergency procedures and operator training to enable a correct response to these problems. Pre-operational recirculation tests performed by utilities have consistently identified the need for plant modifications.
The NRC has begun a generic program to resolve this issue.
However, more immediate actions are required to assure greater reliability of safety system operation. We therefore require you take the following actions to provide additional assurance that long-term cooling of the reactor core can be achieved and maintained following a postulated LOCA.
- 1. Establish a procedure to,perform an inspection of the con-tainment, and the containment sump area in particular, to identify any materials which have the potential for becoming debris capable of blocking the containment sump when required for recirculation of coolant water. Typically, these mate-rials consist. of: plastic bags, step-off pads, health physics instrumentation, welding equipment, scaffolding, metal chips and screws, portable inspection lights, unsecured wood, construction materials and tools as well as other miscellaneous loose equipment. "As licensed" cleanliness should be assured prior to each startup.
September 1981 6A.54-1 Amendment 6 08-03-81
PVNGS FSAR 44c St This inspection shall be performed at the end of each shut-down as soon as practical before containment isolation.
- 2. Institute an inspection program according to the requirements of Regulatory Guide 1.82, Item 14. This item addresses inspection of the containment sump components including screens and intake structures.
Develop and implement procedures for the operator which address both a possible vortexing problem (with consequent pump cavitation) and sump blockage due to debris. These procedures should address all likely scenarios and should list all instrumentation available to the operator {and its location) to aid in detecting problems which may arise, indications the operator should look for, and operator actions to mitigate these problems.
- 4. Pipe breaks, drain flow and channeling of spray flow released below or impinging on the containment water surface in the area of the sump can cause a variety of problems; for example, air entrainment, cavitation and vortex formation.
Describe any changes you plan to make to reduce vortical flow in the neighborhood of the sump. Ideally, flow should approach uniformly from all directions.
l
The following additional guidance is provided for performing this evaluation.
5.1 Refer to the recommendations in Regulatory Guide 1.82 (Section C) which may be of assistance in performing this evaluation.
.5.2 Provide a drawing showing the location of the drain sump relative to containment sumps.
Amendment 6 6A.54-2 08-03-81 September 1981
PVNGS FSAR QAg P 5.3 Provide the following information with your evaluation of debris:
- a. Provide the size of openings in the fine screens and compare this with the minimum dimensions in the pumps which take suction from the sump {or torus), the mini-mum dimension in any spray nozzles and in the fuel assemblies in the reactor core or any other line in the recirculation flow path whose size is comparable to or smaller than the sump screen mesh size in order to show that no flow blockage will occur at. any point past the screen.
- b. Estimate the extent to which debris could block the trash rack or screens {50 percent limit,). If a blockage problem is identified, describe the corrective actions you plan to take (replace insulation, enlarge cages, etc.).
- c. For each type of thermal insulation used in the con-tainment, provide the following information:
(1) type of material including composition and density,
{2) manufacturer and brand name,
{3) method of attachment, (4) location and quantity in containment of each type, (5) an estimate of the tendency of each type to form particles small enough to pass through the fine screen in the suction lines.
- d. Estimate what the effect of these insulation particles would be on the operability and performance of all pumps used for recirculation cooling. Address effects on pump seals and bearings.
RESPONSE
tgR~P~
September 1981 6A.54-3 Amendment 6 08>>03-81
/
~ ~ I
- 1. CESSAR section 16.4.5.2.b commits to inspection of the con-tainment prior to establishing containment integrity.
- 2. CESSAR section 16.4.5.2.c.2 commits to the inspection required by Regulatory Guide 1.82 (Rev. 0) Item 14.
- 3. Fiant procedures will require an operator to
~ ECCS performance during long term recirculation cooling using the ECCS. These procedures will provide specific guidance on recognition and mitigation of ECCS performance degradation during recirculation operation. They will also include guidance to alert the operator to the symptoms of inadequate core cooling. Amended section 6.3.1.4.H.2 refers to CESSAR Table 6.3.2-3 which provides a list of the instrumentation available to the operator to moni-tor ECCS performance.
~,'Et 4.
'There. etre riei high enemy irnes inea4ed in Hie v'(ciniQ nk Ae s~ps uhtc,4 co&cL ~g$ ey~g ~f~
suii<~s'4 ( c pe~ck~ ~M vaguer'rA.
+~~
5.1 The PVNGS design fully meets the requirements of NRC Regulatory Guide 1.82, Revision 0.
5.2 Figure 6A-4 shows the location of the drain sump relative to the containment sump.
5.3.a. Figure 6A-5 provides the size of openings on the screens. No flow blockage will occur beyond the screen as all openings are larger than the minimum screen size.
5.3.b. The estimated+lockage is 20%. The model tests were made for up to<~ I5 /o blockage.
5.3.C(1) Type 304, stainless steel 5.3.c(2) Mirror insulation by Diamond Power Corporation
- 5. 3. e(3 ) Attached by stainless steel buckles 5.3.c(4) Only mirror insulation is used in the containment except for 400 feet of fiberglass insulation used on 10-in., 8-in., and 6-in. chilled water pipe. The fiberglass insulation is manufactured by the CERTAINTEE Company and is surrounded in every application by a stainless steel jacket.
5.3.c(5) The model test of the containment recirculation sump and screen included modeling vario s pe centa es of screen pluggingAC, NRKlhlLC4( SCre85 PllLeIg ana z~iow conaxta.ons.> jhowe e Hmo esNaSte 5t, gerCena.
report describes in detail various test parameters. The report
%+he NRC information has been submitted>as part. of the Containment Systems Independent Design Review submitted under PVNGS transmittal letter'NPP-18147, dated June 4, 1981. This model test report has shown that a vortex breaking cage Amendment 6 6A.54-4 September 1981 08-00-81
PVNGS FSAR GtrL 1 8gisivcl, needs to be installed at the suction pipe. This change will be implemented. No other changes in piping or struc-tures The combination of this testing and the analytical calcula-tions for head loss of piping outside the model's scope, prove that there is adequate NPSH at. the safety injection pumps.
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jJE~E September 1981 Ci ep+'e ~ equi 6A.54-5 08-03-81
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Amendment 6
~,SI DRAIN SUMP RECIRCULATION "A" ECIRCULATION SUMP "B" SUMP
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EXTENT DF MODEL TESTING Palo Verde Nuclear Generating Station FSAR RECIRCULATION SUMPS CONTAINMENT BLDG. EL. 80 6A-4 'igure September 1981 Amendment 6 r
08-03-81
1.1075" 1.25" 21 ~
27 ~
0.1875" 0.25" 4I ROO 0I ~
OUTER GRATING 1.25" x 0.1875" BA R0.5 COARSE SCREEN OUTER GRATING (SCALE: 1" = 2")
FINE INNER SCREEN 3- 75 ~
0 5" 0.5" ~0.12" 77WIRR COARSE SCREEN (ACTUALSIZE) 0.09" 0.09" 0.035" P WIRE FINE INNER SCREEN 27 ~
(SCALE: I" > 0.2")
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Palo Verde Nuclear Generating Station
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FSAR (SCALE: 1' 05 )
CONTAINMENT RECIRCULATION SUMP SCREEN DETAIL Figure 6A-5 September 1981 Amendment 6 08-03-81
PVNGS FSAR APPENDIX 15A QUESTION 1SA.48 (NRC Question 440.82) ~ (15.0}
Section'15D.2.2.2 of the CESSAR System 80 FSAR states that the loss of instrument air event impact on the plant systems and components will. be addressed in the applicant's FSAR.
Discuss the loss of instrument air for Palo Verde showing that, it meets the appropriate acceptance criteria for a moderate frequency event. Causes and potential systems interactions should be addressed and the loss of instrument air should be considered during all phases of reactor operation. Also, present your plans and capability for preoperational or startup tests to substantiate e analyses.
RESPON E: The nitrogen supply system will support the +ufYCtfg instru ent air system for one hour on loss of instrument ts nccomplls'hecL air This <by prdvzdzng an automatic control valve connecting the nitrogen system to the instrument air system. Depletion of the nitrogen system will not affect l any safety related systems.
[QSCR7 5 September 1981 15A.48-1 Amendment 6 07/30/81
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440. 83 (II.B.1)
'four response to Item II.B.l of NUREG-0737 requirements 15 sufficient. Provide the following:
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- 1. Provide diagrams and a description of the vent discharge vicinity. Verify that adequate ventilation is provided and that equipment in this area is capable of withstanding discharge of gases and liquids from the vents'.
What size are the flow limiting orifices and what are the calculated flow rates through the vent system for both gas mixtures and liquids at operating pressures'?
- 3. Provide drawings of the piping system from the vessel E
head and pressurizer through the discharge paths. In
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particular, show the location of the solenoid operated jo valves and consider potential I'.ssile hazards from them.
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eclr> ~ ~p(eesmkegha gsS Ld" l ~ oiqure ~.,~-) eeg~oQ cot[i $ e Line RC-148-BCBA-1" discharges into an open area near steam provide/ fN'it 66cre L~IR otoeivgjk@p generator number 1. This area is not restricted in any way. This occurs at elvation 158'6" at the north side of the containment.
There is adequate ventilation and there is no~wtipment in the area of the discharge that could be affected by system operation.
2.The flow limiting orifices have a round opening of 7/32".
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Anticipated flow rate is about 500 scfm.
- 3. The system was reviewed on the plant>model found that there acedia no credible missile@>
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To minimize the possib'ility of common mode failure of solenoid operated valves to shut when de-energized, the operation procedure for the Reactor Coolant Gas Vent System (RCGVS) will require that Q.vc clvai (a.b ~C when ~ both Trains A and B that one valve powered from Train A and one valve powered from Train 3 will be used to complete a vent path.
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QueSk'f777 440e84 Your response nf ~~ ~
to Item II.K.3e17; of NUREG0737 is not com-plete. Provide a commitment that you will establish a program prior to fuel loading for data collection on information regarding ECCS outages. The information will.contain: (1) outage dates and duration of outages; (2) cau'se of the outages; (3) EECS systems or components involved in the outage; and (4) collective action taken.
PtllCS Response: g@aeended Section II.K.3.17 of LLIR.
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..S,vera). corn'pnnents of the emergelicy core cooling {I:.CC) sy.:I.
are permit.l.ed key'echnical speci ficatio>>s I.o have .".>>t:.sI a>>~ I times {e.g., 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> for n>>e dies.l-qe>>c-.'rator;
.'>>tage 1.4 for the HPCI system). In addition, t)iere axe no c>>mulati:r outage time limitations for ).CC system . I,icen."ee.". s)>o>>l()
s>>bmit a report'detailing outage dates alld 'lengths of outag""
for all ECC systems for the last 5 years of nperation. T)i~
report sl>q>>ld also incl>>de the caus~s of th o>> tages { e. g.,
control] er fai lures, spurious isola tioll) .
P~J1'IGS Evaluation prop~~
>>y>>'ll t s>eblis), c collection of oi)tage dates and Iellgths of oui:age..:>~
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1 440. 85 'our response to Item II. K.3.25 of HUREG-0737 states that the (I I. K.3. 25) ,reactor coolant pump normal cooling water system (nonsafety Palo Verde 'rade nuclear poolina water system) is backed up by the only , essential seals during loss of offsite AC power. Describe the manual action involved agf tba ms~ g~ limo zam~s>d.g,o.r transferrino the cooliSg, ~>> ".mpaasas . Also, state that your operating procedure allows enough time to restore the cooling water supplies to the RCP seals before you trip the RCPs. After the RCP trip, you may still need essential cooling water supply to the RCP seals.
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PIGS FSAR tt"" '"g (6.3)
Expand your interface requirements in Section 6.3.1.3 to include the requirement of power locked out on the SIS hotleg injection valves in order to prevent premature hotleg injection following a LOCA.
RESPONSE: The response will be provided on the CESSAR docket. Zjn add<QGkf> a.X S~n on FSAR Fi'uYY Q Q j >
khQ hQ)~ f4jtgiOA f50lghoA vklv8s -(Sx'A- Hv(gogy 3~ jg see-HVGN~ Ml ) are postern! ~vn zePara4e ~eel 5~)PheS.
September 1981 6A.55-1 Amendment 6 07"30-83.
Q(eskb~ 440. F7 lt l
~.g7 Q"""""q (5.41 anh "9.2.2)
If the RCP tests demonstrate that the RCPs are capable to oper.-
ate with loss of component cooling water supply for longer than 30 minutes without loss of function and the need for operator protective action, safety grade instrumentation to detect the loss of component cooling water to the RCPs and to alarm the operator in the control room should be provided.
The entire instrumentation system, including audible and visual status indicators for loss of component cooling water should
.eet the requirements of IEEE std. 279-1971/1974. The 'above requirements should be specified in the applicable section (e.g., Section 5.4.1 or 9.2.2) of CESSAR System 80 FSAR as interface requirements.
RESPONSE: edundant Class IE flow transmitters are provided
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for each nuclear cooling water supply to the RCP+Coolers.
This instrumentation provides visual and audible annuncia-tion to the control room'operator on loss of nuclear cooling water flow. The redundant Class IE annunciators are dis-cussed in section 7.6 and meet the requirements of IEEE Standard 279-1971.
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