ML20028B809
| ML20028B809 | |
| Person / Time | |
|---|---|
| Site: | Washington Public Power Supply System |
| Issue date: | 11/26/1982 |
| From: | Novak T Office of Nuclear Reactor Regulation |
| To: | Ferguson R WASHINGTON PUBLIC POWER SUPPLY SYSTEM |
| References | |
| NUDOCS 8212060341 | |
| Download: ML20028B809 (18) | |
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g NOV 26 haz DISTRIBUTION sOschet llo. 50-460L LB#4 r/f -
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NRC PDR Uncket fio.
50-460 L PDR PRC System NSIC EAdensam Pr. R. L. Ferguson MThadani Managinn Director MDuncan L'ashington Public Power Supply Systen DEisenhut/RPurple P.O. Box 968 Attorney, OELD 3000 George Washington Way ELJordan,DEQA:I&E Richland, WA 99352 JMTaylor, DRP:I&E
Dear Mr. Ferguson:
Sub.fect: Request for Additional Information on Washington Nuclear Project, Unit No. 1 In our letter dated July 16, 1982, regarding the acceptance of your application for operating license for liashington Nuclear Plant Pro.fect, Unit Ho 1, it was indicated to you that it is the staff's intent to proceed on a "nanpower available" basis with review of those portions of the application which parallel other current applications of sinilar design or with similar features.
In accordance with this intent the staff is eviewing the appropriate portions of the Final Safety Analysis Report (TSAR) for Washington fluclear Project, Unit 1, and is in the process of developing input for the Safety Evaluation Report (SER).
In the course of this reviev, the Auxiliary Systems Rranch has identified, in the enclosure, additional inforration necessary in order to complete the revieu and prepare an input to the Safety Evaluation Report.
Please supply the additional infornation within 30 days of the receipt of this letter. Should you have any questions on tha attached, contact ifr. Mohan Thadani at (301) 492-8941.
The reporting and/or recordkeeping requirenents contained in this letter affect fewer than ten respondents; therefore, OtM clearance is not required under P.L.96-511.
Sincerely, 8212060341 821126 PDR ADOCK 05000460 Thonas H. Novak, Assistant Director E
PDR for Licensing Division of Licensing
Enclosure:
As stated
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OFFICIAL RECORD COPY usomini-- m m unc ronu aia 0o-80 Nncu cao
WNP Mr. R. L. Ferguson "
Managing Director Washincton Public Power Supply System P.O. Box 968 3000 George Washington Way Richl and, Washington 99352 cc:
Mr. V. Mani Mr. George Hanson United Engineers & Constructors, Inc.
State'of Washington 30 South 17th Street Energy Faciltiy Site Evaluation Philadelphia, Pennsylvania 19101 Council
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Mail Stop PY-11 Nicholas S. Reynolds, Esq.
Olympia, Washington 98504 Debevoise & Liberman 1200 Seventeenth Street, N.W., Suite 700 Washington, D. C.
20036 Mr. E. G. Ward Senior Project Manager Babcock & Wilcox Company P.O. Box 1260 Lynchourg, Virginia 23505 i
Resident Inspector /WPPSS NPS QU.S. Nuclear Regulatory Commission P.O. Box 69 Richland, Washington 99352 Mr. R. B. Borsum Nuclear Power Generation Division
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Pcbccck & Wilcox 79:0 Woodmont Avenue, Suite 220 Bethesda, Maryland 20814 G. E. Craig Doupe, Esq.
Washington Public Power Supply System 3000 George Washington Way P.O. Box 968 Richland, Washington 99352 Robert Engelken, Regional Administrator U.S. Nuclear Regulatory Commission, Region V 1450 Maria Lane, Suite 210 Walnut Creek, California 94596 O
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1 AUXILIARY SYSTEMS ERANCH REQUEST FOR ADDITIONAL INFORMATION WASHINGTON NUCLEAR PROJECT NO. 1 DOCKET NO. 50-460 410.1 Are the ccmcartments containing safety related equipment watertight?
(3.".1)
Are the interior doors to these compartments watertight? Are the well penetrations (both electrical and rechanical) watertight?
What is the cesign pressure for these water seals?
410.2 In tne analysis for internal flooding, no operator action was assumed (3.4.1) for l'0 minutes.
Provide the results of a similar analysis which includes the following additional assumptions:
- No operator action for 20 minutes in the control room, 30 minutes if required outside of the control room.
- All non-seismic Category I piping, tanks, sumps, valves, and equip-ment fail.
Frcvide the minimum flood level required to affect each safety related ccmponent.
41 0. 3 Verify that high pressure gas bottles and accumulators were considered
( 3. 5.1 )
as potential missiles in your analysis of internally generated missiles
. impacting on safety related equipment (both inside and outside contain-pctential missile sour'ces into consic, cuss how you will take thes ent).
If this was not.the case, diseration in the design of the plant.
41 0. 4 Verify that your analysis included the generating or impinging of
- (3. 5.1 )
internally generated missiles on safety related equipment required to achieve and aintain cold shutdown.
If this was not the' case, discuss tr.e procedure f or acnieving and maintaining ccid shutdenn for ea'ch identified internally generated missile assuming the failure of.all cold shutoown equipment in the path and range of the missile.
41 0.5 Verify that any internally generated missile from safety related
' 3. E.l }
equipment will not affect the redundant safety related train.
210.5 Verify that the following potential missile sources 'inside containment
( 3. 5.1 )
have been included in your evaluation and that safety related equipment has tean ;-otected..
1.
Reactor vessel a.
closure head nut b.
incere instrumentation assembly O
1 o 2.
primary manway stud and nut b.
secondary handhole tud and nut c.
se:cndary manway stud and nut 3.
Pressurizer a.
safety valve with flange b.
safety valve flange bolt c.
relief valve with flange d.
safety valve from bonnet flange e.
lower temperature element f.
manway s;ud and nut i
4.
Main coolant piping temperature nozzle with resistance temperature detector i
5.
Surge and spray pipi,ng wells witr. resistance temperature detector assembly 6.
P.eactor coolant pump thertowell with resistance temperature detector 7.
Shutdcwn ccoling valve stem 8.
Reactor coolant pump mounting flange. leakoff connections 410.7 The general arrangement drawings do not indicate tornado missile pro-
{2.E.2) te:tior. #:r.ecuip ent listed below.
Provide detailed drawings of the tornado missile protection for each of_ the following:
- diesel cenerator exhausts
- atmospheric dump valve exhausts
- safety relief valve exhau^ts
- every heating and ventilating system intake and exhadst 4
O h
9
. - remote air intakes for the centrol room ventilation system
- spray pends
- spray trees
/n.2 Tne fi;ures ;isted in Table 3d-2 are not in tne FSq.R.
Provide drawings
( 3. 5.1 )
for ali nigh energy.iines outside of containmer. whicn include asi pipe anchors, break locations, and high stress locations.
The construction permit application was tendered on July 16, 1973.
410.9 The FSAR states $ ranch Technical Position ASB 3-1. conta t
(3.6.1)
Based on the date the construction permit app'lication was tendered for WNP-1, BTP ASB 3-1 Paragraph B.4.b requires the applicant to conforin to the July 12, 1973 letter from J. F. O' Leary or the position itself.
It is not clear in the FSAR which option was selected for WNP-1.
Provide a statement of
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coriformance to either the J. F. O' Leary letter or the Branch Technical Position itself.
410.10 The FSAR states that the cooling water for the control rod drive (4.6) mechanisms (CRDMs) is not required to maintain thc ability :to trip.
With the reactor at full pcwer, how long can the CRDPs be without
" cooling water before the insertion time of the control rod assemblies is affected?
Is the less of cooling water to a CRDM alarmed in the control room?
If the time required for operator action is twenty minutes or less (30 minutes if the recuired action is outside of the control reon) from the time of the alarm, provide a discussion of the operating procedure to bring the plant to cold shutdown.
Provide drah.;s of CR:"s shcwing the details of the mechanisms and layout drawings showing the routing of the cooling water piping and the electrical cables up to and including the containment penetrations.
410.11 All control rods need to be tested monthly and after each ~ refueling.
(4.6)
Discuss the procedures used to perform each test.
i 410.12 Verify that the soluble poison cor. trol is capable of' maintaining the (4. 6) core subtritical under conditicns of cold shutdcwn independent of the l
410.13 In measuring the identified ieakage rate, the FSAR states there are (5.2.5) two level sensors that monitor the water flow through a weir.
One sensor is connected to the plant computer and the second sensor is connected to a level indicator on the MSS Primary Panel.1 An annun-l ciator is activated when the 1.0 gom flow rate is exceeded.
Verify that the level indicator on the MSS Primary Panel activates the annunciator.
O f
4_
-10.14 The FSAR states that each raci ; articulate and radiogas monitor pro-
'E.2.E; vides a sigr.ai to two sets of electronics. One set of electronics provices an cutput in,4Ci/cc and the other set uses a digital dis-criminator set at 2 MEV for radioparticulate and 1 MEV for radiogas moni:0 ring.
The FSAR is n:t clear as to what the sensitivity for the moni crs and, ultimately, the readouts are.
V-erify that the raci:;a-ti:ulate and the radic;a_s instrumentation detect and indicate a radicaut: city of 10 Y and 10-t//C1/cc, respectiveiv.
ai 0.15 Describe what the " optimum credible moderation" condition is.
How is
( 9.1.1 )
this condition different from optimum moderation? What is the Keff for the new fuel storage facility under the optimum moderation conditions?
410.16 The discussion provided in the FSAR is not clear with respect to the
( 9.1.1 )
design of the new fuel storage racks.
Provide a detailed drawing of the fuel cell connections to the fuel rack structure and the dry empty weight of a fuel cel..
Verify that an uplift force equal to the dry l
empty weight of a fuel cell will cause the failure of the attachments of the cell to the structure.
Assuming the fuel cell is removed from the rack v, hen a fuel assembly is being removed, discuss the actions c be taken by the operator and how the rack would be repaired.
1 410.17 Verify conformance with the ANS 57.2 standard, " Design Objectives (9.1.2) for Light Water Reactor Spent Fuel Storage Facilities at Nuclear Power ea> Stations," as specified by NUREG-OS00, Standard Review Plan, Section 9.1. 2.
410.15 Verify that the spent fuel pool lir.er and all gates which separate (9.1.2) the spent fuel pool from the cask loading area and the transfer canal are seismic Category I.
Prc ei:e a discussion ar.d drawings of the spent fuel p;o1 liner l'eakage 4 0. '. 9
( g.1. 2 )
detection system.
410.20 In order to facilitate indeper. der.t evalua' tion of the criticality of (9.1.2) the spent fuel storage racks please provide the folicwing additional data:
1.
The inner or outer (specify which) dim'ension of the two stainless steel tubes, 2.
The s:tinless steel type, 3.
The nominal values of the density of boron carbide and carbon in the poison slabs and the uncertainties therein.
i 1
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41 0.21 In order Oc facilitate evaluation of the computation methods used
':.1.2)
- iease provide the feilcwing additional information concerning methods verification
1.
Description of the ex;erirents acainst which verification was performed witn particular emphasis on experimeots which contained features relevant to stcra;e rack design (e.g., poison slabs between assemblies, water ;aps re:-een assenblies, etc.).
Provide infor-mation for botn che KEND and PDQ code packages.
2.
Indicate the extent of the verification performed by the applicant or his contractor (apparently NUS) and that performed by other users.
Greatest weight will be given to the fomer.
410.22 The applicant has not provided an analysis which confoms to the (9.1.3)
Standard Review Plan (NUREG-0800) Section 9.1.3 with respect to the RSP cooling time of 150 hours0.00174 days <br />0.0417 hours <br />2.480159e-4 weeks <br />5.7075e-5 months <br /> for the last batch into spent fuel storage, the. definitions of norgal and maximum normal heat loads, and the maxi-The applicant num temperature of 140 F with a single. active failure.
used only la reloads when 16 relcad can be placed in the spent ft;el s:Orage racks based en the eccilibrium fuel cycle reload of 35 fuel assemblies indicated in F5A. Table 9.i-2.
The applicant is to prcvide a discussion of this scenario, the revised calculations, and a revised FSAR section.
410.23 q>,All calculations of the decay heat loads shall be in accordance with
( 9.1. 2) the Branen Technical ?csi ion ASB 9-2, "Residuel Decay Energy for ligh: Water Reactors for ' cn; Term Cooling." Provide the results of revised calculations using the Branch Technical Position.
'410.24 Verify that direct indicatter. of the spent fuel pool temperature and 9.'.?:
level is available tc the c:e-ator in the centrol roen.
The FSAR states that the low ficw alarm in the spent fuel pool c'ooling 410.25
' 9.1. 3 )
system is disabled when the associated cooling pumn is off.
Provide a -list of alarms which wcaid define a loss of a cooling pump to the operator.
410.26 Describe, discuss, ar.d verify that the maximum pctential kinetic energy (9.1.4) centained in all objects of less weight than a spent + fuel assembly which will be handle: over spent fuel in the storage racks will not exceed the effects c' the fuel handling accident in Section 15.7.4 of the FIAR.
Verify compliance with the guidelines of ANS 57.2-1976.
For each item 410.27 (9.1.4) where the guidelines are not met, identify the item and provi.de a dis-(9.1.5) cussion of the deviation.
t Verify that the fuel transfer tube gate valve is seismic Category I.
410.28 (9.1.4) e
O i
a 410.29 Provice tne.information recuested in the generic letter dated December 22,
( 9.1. 5) 1920 recarding conformance to the criteria contained in NUREG-0612, "Controlf oi Heavy Loads at Nuclear Power Plants."
212.30 Provide a list of all load handling systemsand identify which systems
'?.l.4) are used in moving loads weighing mere than one fcel assembly and its asscciated handling tool.
In addition, identify those systems which can ncve any 1.cac cver sper.t fuei either in the storage pool or the cpen reactor vessel.
410.31 Provide an explanation of why nuclear service water valves are in the (9.2.1) component cooling water system as shown in the FSAR Figure 9.2-11.
410.32 FSAR Figure 9.2.-1A indicat,es that valve MSW-V75-A is seismic Category I (9.2.1) while FSAR Figure 9.2-11 indicates that this valve is non-seismic.
Category I.
Provide a clarification as to whether the valve is seismic Category I or not.
E10.33 FSAk Tables 9.2-3 and 9.2-4 are not clear.
Discuss how the shutdown (9.2.1) ccoling water system's (SCWS) heat lead is 103.91 MBTI'/hr (Table 9.2-4) wnile the nuclear service water system removes 214.88 MBTU/hr (Table 9.2.-3) frcm the SCWS, on a per train basis.
410.34
- According to the FSAR, there are two redundant nuclear service water
( 9. 2.1 )
- > systems and either system can remove 100% of the plant heat load.
Table 9.2-1 in the FSAR specifies a meximum flow requirement for one train wi n a minimum flow requirement for the other train.
Provide an explanation for this discrepancy.
.she FSAR states that the shutdown cooling water system is required to 410.33
'?.2.1) supply rater to the spent fuel pool heat exchanger "a minimum of 1.75; hcurs" af ter the icss of the conponent cccling water system.
This statement is ambiguous.
Discuss why the spent fuel pool heat exchanger can operate a " minimum of 1.75 hours8.680556e-4 days <br />0.0208 hours <br />1.240079e-4 weeks <br />2.85375e-5 months <br />" without cooling.
Include the conditions which are used to cefine this minimum, i
Frovide a discussicn cf the inservice inspection and testing program
~
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- 1~.35 l
(9.2.1) for the nuclear service water and shutdown. cooling water systems.
Included should be the identificatien of valves to be monitored, and what functions are performed from the centrol room and which are local occ ations.
Is scund ;cwered ccm unication with the control recm iccalij available for the cperators at the testing station to instruct the test personnel to place the equipment back in service if the equipment was required to be inservice?
n 41 0.37 The FSAR provides the following information.
Table 9.2-5 states that (9.2.2) the shutdcun cooling water pumps have a design flow of 13,500 gpm Section 9. 2.1. 2. 2. 4 states that only one shutdown cooling water each.
system can be cross-connected to supply the essential equipment in O
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the component cooli.ng water system at at given time.
Furthermcre, i
this section goes on to specify that if the-13,500 spm flow is avail-able, 7,500 gpm is provided to the decay heat re oval heat exchanger, 4,500 gpm is provided to the containment spray heat exchanger, and
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the remair. der (1,500 gpm) is provided to the spent fuel pool heat excharger.
The decay heat removal and containment spray heat exchan-pers are part of the normal shutdcwn ecoling system.
The spent fLei pool heat exchanger is part of the component cooling water system.
~
Table 9.2-8 indicates a required flow of 4,905 gpm through the cross-connsction during a loss of offsite power event.
Considering the loss of the redundant shutdown cooling system as the single failure
~
and the reyised decay heat load calculation (Question 410.23), prc-vide an explanation of the apparent discrepancy in the required vs'..
available flow through the ' cross-connection.
410.38 FSAR Table 9.2-8 indicates lass cooling water flow requirements for (9.2.2) the essential safety related equipment after a LOCA concur. rent ' ith w
a safe shutdown earthquake and the resulting loss of offsite power than required during a loss of offsite power.
Provide the information which indicates your conclusion.
For example, with the maxinum normal spent fuel d.ecay heat load as recalculated in Question 410.23 and main-taining a maximum pool temperature of 1540F as per the FSAR, discu'ss why a loss of offsite power requires 2,100 to 3,594 gpm coolant flow.
,as -compared to the 1,750 gpn required during a LOCA, plus SSE and the resulting loss of offsite power.
\\
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~410. 39 Describe the component operational degradation (e.g., pump leakage)
(g.2.2) ar.d the procedu.res that will be followed to detect and correct these conditions when degradation becomes excessive.
410.4; In FEA; Table 9."-9 the teral heat lead on the cc ponent cccling water (9.2.2) system during normal plant operation is 127.94 METU/hr per cooli.ng loop while Table's 9.2-3 and 9.2-10 specify a cooling capacity for the cceponent cooling water heat exchanger of 57 MBTU/hr.
Provjde a discussion which explair.: the apparent discrepancy.
410.41 Discuss the effects on the spray nozzles and piping,of pumping 23,000 (9.2.3) gpm fron the emergency shutdown water system when the spray system is designed for 21,465 gpm, which represents 107.15% of the spray syste-design ficw..
410.42 The FSAR states that the balance of plant service wat'er system consists (9.2.4) of three pumps.
Each pump is rated for 50% of the normal full load cooling water flow requirements for a flow of 20,000 gpm each, as per Table 9.2-18.
FSAR Table 9.2-17 indicates a required normal full load cooling water flow requirement of 51,645 com total.
Provide a discussion of this apparent discrepancy.
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In FSAR Tacle 9.2-24 the total integrated heat load to the ultimate t
3./.5) heat sink following 4 loss of effsite powe" is 34,048 MBTU/hr.
, Tatie.S.\\2-23 indicates a heat load of apprcximately 58,000 MBTU/hr.
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'F avice a discussicn of this acparent ciscrepancy..
..d':."
In :te 253. Table 9.2-25 the heat load to the ultimate heat sink is
- ;. 2. 5 )
for a loss of coolant and loss of o.ffsite pcwer accicent.
The rate 1 of heat rejection to the ultimate heat sink and the total heat s
' rejected indicated in this table is less than in Table 9.2-23 which is only for a loss of offsite power.
The heat. load indicated in iable 9.2-25 does not agree with the heat load to the ultimate heat sink indicated in Table 9.2-24 (this. table is applicable by reference from Table 9.2-26).
Provide a discussion of this apparent discrepancy.
410.45 Table 9.2-92 in the FSAR indicates that the loss of one. spray system
~
(9.2.5)'
.in the ultimate heat sink due to a to nado will have no effect on the RSP cooling capability of the pond because of the redundant spray tree syste :.
Unless the sDray trees are tornado missile p'rotected, no credit can be taken for any; spray tree system or portion of a spray tree syste..
Provide the results of an analysis as the result of not due to tornado missiles, and the most limiting having any spray trees., As an alternate, provide complete tornado sincie active failure.
missile protection for the spray tree systems.
2: 0.45' Eeguia ory Guide 1.27 recuires khat there be sufficient water in the (9.2.5) ipray p;nds for cooling without makeup.
Discuss how you will monitor the buildup of sediment on the floor of the ponds so as to assure a#'t.ilability of the 30-day water supply.
Describe how you will clean the spray ponds withcut losing redundancy or ' degrading the system.
' the syste-410.47 Since 'n'h?-4 has been withdrawn, discuss the effect on the equipmer.t (3.2.5)
.and design of the demineralized water makeup, potable water and sani tar;. yystems at 'n'NP-1.
- 0.43 5pecify the seismic categorization of the water treatment building.
(9.2.6)
If the building is not seismic Category I*, $rovide a< discussion of the effect of the collapse of the building and the systems within the building on the main steam and feedwater lir.es ar.d the main stear ar.c feecnater isolaticr. area whi:h shares a cc=:n wall with the water tre m ent building.
410.49 For the Plant Service Air System, the Instrument Air Supply System, (9.3.1) and the Nuclear Instrument Air System, provide the follow.ing:
1.
A commitment to perform periodic testing of the air system.
l 2.
Verification of conformance with Regulatory Guide 1.80, " Pre-operational Testing of Instrument Air Systems."
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3.
Verification that the air system will provide an air quality as specified in ANSI MC 11.1-1976 (ISA 57.3) or better.
FSAR Table 9.3-5 indicates non-compliance with ANSI MC 11.1-1976.
410.50 Portions of the non-seismic Category I compressed air systems are (E.2.1) ic:sted insice the General Service Euilding which is seismic Category I and contains safety related equipment.
Verify that the complete failure of the non-seismic Category I compressed air system and its supports will not affect any safety related equipment as the result of a safe shutdown earthquake. As part of your verification, provide general arrangement drawings which show the safety-related equipment and the seismic and non-seismic Category I compressed air piping.
410.51 Figures 9.3-1 and 9.3-2 in the FSAR do not identify dhere the comoresud (9.3.1) air. is used.
Provide a revised set of drawings which clearly iden-tifies the equipment or component being supplied compressed air.
If your response to Question,410.50 shows all ccmpressed air piping,
- en -his :uestien need nct be addressed.
410.52 The FSAR states that the instrumert air supply system (IAS) supplies (9.3.1) the nuclear instrument air system (IAC) under nomal operating con-Qi tions.
Furthermore, the IAS maintains a header pressure between 0 and 100 psig.
Provide a discussicn which explains how the IAC neader pressure is to be maintained a: 100 to 125 psig under nomal operating conditions as indicated in the FSAR.
41 0.53 Provide a full si:e riser diagram the radioactive equipment and (9.3.3) ficor drainage piping system which identifies each potential water a rce, valve:, sumps, and intercennections.
Prcvide a sinilar
- diacram.tne non-radioactive equipment and floor drainage piping system.
For each area where there is both radioactive and nonradio-active drains or piping, provide an isometric layout drawing und a discussion of any event where radioactive fluid could enter the non-radioactive drain as the result of excessive flow, plugage of the radioactive drain, or pipe failure.
410.54 Verify that the containment isolation val *ves V-22-B,< V-23-B, V-72-A
- v. :.1) anc '.'-73-5 are seismi.c Ca:egcrj I with a Cla:s IE pcter supply.
410.55 Provide a pnysical drawing of the control rccm remote air intakes (9.4.1) and a location drawing.
Verify that they are seismic.Categcry I and tornado missile protected.
410.56 The FSAR states that the electric heaters in the control troorn HVAC (9.4.1) are not Class 1E.
In some design basis events, such as an Appendix R fire, offsite power is assumed lest for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. Assuming such an event occurred durinn the most severe winter weather, describe how the control rocm will be maintained at750F DB and 45-50% RH.
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- 1 0-
"10.57 The FSAR Tigure 9.4-5 indicates that there are two dampers in series
- 9.".1) for each, intake of the remote air in
- akes.
One damper is supplied power from one Class 1E bus and the second damper is odwered by the redundan: Class lE bus.
Therefore i appears that.the failure of one Class iE bus will isolate both remote air intakes.
If this indeed is corre::, it is unacceptable.
Provide a discussion ar.d a revised drawing to clarify the arrangement of the dampers and their power sources.
410.58 Assuming the need to use the remote air intake and a high radiation (9.4.1) level in one of the air intakes, describe how that air intake will be purged until the radiation level is within acceptable limits without supplying air to the rest of the control room HVAC system.
410.59 Provide one or more sections in the FSAR which describe all of the
( 9. 4.1 )
chilled water systems with P& ids, layout drawings, and seismic qualifications.
ii.f h The FSAR is not ciear concerning the continuous minimum flow of 4,000 r'e CFM exnaus: from the fuel handiing area ventilation system. Verify compliance with Position C.4 of Regulatory Guide 1.13 and the use of Regulatory Guide 1.25 as the minimum potential source of radiation
,,(pr a fuel handling accident.
If the number of fuel pins damaged by the accident evaluated in Question 410.26 is greater than the design basis case specified in Regulatory Guide 1.25, then the design require-ments in Regulatory Guide 1.25 should be increased to take into con-sideration the larger number of the fuel pins damaged as the design requirement for the fuel handling area ventilation.
"10.61 Provide the results of an evaluation of the environmental conditions vs.
(9.4.5) equipment qualification for safety related equipment serviced by the safeguards area ventilation system for each of the following conditions.
1.
A safe shutdown earthquake, with the resulting loss of offsite power, during the design -100F winter day with the single active fail re of one diesel resulting in the loss of cne HVAC train.
l 2.
A ::r.adic ever.t, with '5e -c-"l:ir.; less of cffsite pcwer, during the design 110CF summer day with the single active failure j
of one diesel resulting in the less of or.e HVAC train.
0
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410.52 The FSAR states that when the temperature exceeds 104 F, the temperature
( 9. 4. 5 )
controller opens a valve to provide the safeguards area ventilation system with chilled water from the chillers for the spents fuel pool area ventilation system. A reference is made to Section 9.4.2 of the FSAR for details.
There appears to be no mention in FSAR Section 9.4.2 of this ability and FSAR Figures 9.4-1 and 9.4-7 do not show any such capabili ty.
Provide the details which were to be in FSAR Section 9.4.2 and a drawing which shows every HVAC system with all interconnections to support systems and equipment.
. *10.63 The FSAR states that the air is drawn by four fans into the control (9. 4. 6) rod drive area.
This does not agree with FSAR Figure 9.4-8.
Provide aither 'a revised description or a revised figure.
il0.64 The FSAR states tnat the Primary Auxiliary Area HVAC does not operate
- 9.4.7) during a LOCA/ LOOP and that safety related equipmen needed during a LOCA/ LOOP is serviced by this HVAC system.
The FSAC. states that sufficient air flow will be provided by manually opening the manual FSAR Figure 9.4-10 does not show this safety related equip-damp.ers.
ment, the manual dampers to be opened during a LOCA/ LOOP, and the Provide a revised figure emergency air flow via the manual dampers.
whuch provides this infomation.
Assuming a LOCA/ LOOP, 30 minutes after the LOCA befgre the manual dampers are opened, and an ambient temperature of 110 F, what' is the maximum ambient tempera.ture in the room with safety related equipment? To what temperature is this Are the manual dampers remote-equipment environmentally qualified.Are the manual dampers remote-manual.
manually qualified?
If not, verify that the operator can locally open all the manual dampers without passing through any area in which the local environ-ment is being affected by the LOCA.
410.65 Verify that FSAR Figure 9.4-10 is correct with respect to air flow into and out of each area.
For example, the two R.B.C. Evaporation (9.4.7)
- -Surface Condenser rooms have 5,800 cfm and 4,000 cfm flowing into the recms and neither room bas any exhau,st air flow.
410.66 The FSAR states that the evaporative coolers in the diesel generator
~
(9.4.8)
HVAC are not seismic Category I because cooling is not required during a LOCA/ LOOP cr other emergencies.
Asscaing a safe shutdown earthquake, the resulting LOOP, and an amhier.t temperature of 1100F, verify, that
- ne roca temperature for the diesei generators and associated ecuipment I
will be maintained below the equipment qualification temperature or i
1300F, which ever is lower.
Specify the calculated maximum reca l
temperatures.
The FSAR ctates that the switchgear, battery, and cable spreading room 410.67 l
(9.4.9)
HVAC system "does not provide life secooYt or removal of radioactivity in the area." Where will the life support ecu;ccent and radiation pro-tective garnents be kept for cperatcr use in the event of an Apoendix
. fire in the contr:1 rocm? Assuming an Aopendix R fire in the con-trol room, verify that no operator action is required for a minimum of 45 minutes in order for the operator to scram the' reactor, don radiation protective clothing and air packs prior to entry into the switchgear room.
41 0.58 The FSAR specifies portable radiation monitors for monitoring and (9.4.12) alarTaing high radiation level.
The radiation monitoring should be fixed with a high radiation annunciation in the control room.
Verify that the radiation monitoring will incorporate these features.
9
s i" 6:
The FSAR states there is one axial vane return air fan per air handling
'b.52) unit (AMU) train.
This is not consistent with FSAR Figure 9.4-3 which shows the two return air fans having a co : on suction header and common discharte header.
The discharge header feeds a single duct whien later
~
civides into three ducts, one for each AHU and one exhaust duct. There-fcre either return air fan can feed either AHU.
Provide a revised system description or a revised figure, as appropriate.
410.70 Provide a discussion of the design provisions that permit appropriate
( 9. 4.12) inservice inspection and testing of the electrical and piping tunnel HVAC system components and a description of the inservice inspection and testing program.
410.71 Verify that all components in the general services building which a're (9.4.13) served by the component cooling and auxiliary pump room HVAC system are qualified and can operate without any degradation in performance for a minimum of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> in,an environment of 1300F 'and 100% relative humidity.
This includes the HVAC system itself.
41C.72 Wi:n respect to the main steam and feedwater isolation area HVAC system, (9.4.14) the FSAR specifies the ductwork supports and hangers are designed to seismic Category I requirements in Section 9.4.14.1 and also specifies g ne ductwerk supports and tangers are designed to non-seismic require-ments in Section 9.4.14.3.
Provide a revision to the approoriate section.
If the ductwork support and' hangers are not seismic Category I provide a discussion of the effects of this k falling onto the isolation valves due to a safe shutdown eat
'ts associated loss o,f offsite power and the most limiting sing).
=ilure.
C :.73 Provide a description of the inservice inspection and testing pro' gram ~
(9.4.15) for the spray pond pump house HVAC system.
410.74 Verify that all safety related components in the spray pond pumphouse ~
( 9. 4.15 )
are qualifiea and can operate without any degradation in performance for a minimum for 72' hours in an environment of 121oF and 100% relative humidity.
410.75 In accordance with FSAR Section 9.4.12, the safety related portion of (9..15) tne ele::rical and piping tenr.el ventilation system is powered by tne Vitai F:ner ;eteorK (VPO.
The non-safety reia:ed pcrtion nf this system is pcwered by the Auxiliary Power Network -(AH;).
Verify that the seismic Category I Mechanical Equipment Room HVAC, which is safety-related, is powered by the APN as indicated in FSAR Section 9.4.16.3.1.
If this safety related system is powered by the AFN, provide a discussion and drawings which explain the interconnection between the Class lE portion of the APN aad the non-Class lE portion of the APN.
In addition, provide a list of all HVAC components which are powered by the APN as to which are powered from the Class lE APN and which are powered from the non-Class 1E APN.
a-.----
. 410.76 Provide a discussion of the design provisions that permit appropriate
~
(9. 4.16) inservice insp'ection and testing of the mechanical equipment roca HVAC system components and a description of the inservice inspection and testing program.
210.77 The drawir.;r proviced in the FSA?, such as Ficure 9.4-19, do not (9.4.17) adecuately show the.tcrna:o missile protection for the ventila icn intakes.
It appears that the missile protection is not high enough to pr. event entrance of tornado missiles.
Provide physical drawings of each air intake and exhaust which shows the exact placement of missile shields, ducts, openings, and materials of construction.
415.78 The FSAR states that there are pneumatic operators.to manually reopen
( 9. 4.17 )
the tornado valves.
Verify'that the air supply to these valves is from the Nuclear Instrument Air System.
Acdording to FSAR Figure 9.4-20, the exhaust ducts for the ventilation 410.79 systems which service the control, switchgear, battery, and cable (9.4.17) spreading rooms, the electrical and piping tunnels, nuclear instrument air system, makeup (charging) system, and the component cooling system all pass through a single duct chase with fire dampers at each wall Assuming an Appendix R fire in the duct chase, which penetration.
results in closure of all these exhaust systems, during a summer day
- with the ambient temperature of 110 F, provide a discussion of the 0
effects on the ability to safely control the plant.
According to the FSAR figures, there appears to be only one radiation 410.80 nonitor for all. of the exhaust systems.
Verify this is correct.
If (9.4.17) there is only one radiation monitor, discuss how this meets the single failure criterion. If there is more than one radiation nonitor,, provide revisec crawing;.
According to the FSAR figures, the radiation' monitoring of the exhausts 410.81
.s performed only in the exhaust plenum.
Provide a discuss. ion of the (9.4.5/j information available to the operator in the control room and of-the procedcre the cperatcr is to follcw upon receiot of a high radiation alarm frce the exhaust plenum.
Include in this disc,ussion, information on how the exhaust plenum is isolated, air is rerouted for processing prior to being released, and for each ventilation system exhaust duct 50w the :peratcr w '-' identify which system. is releasing radioactivity.
d According to the FSAR figures, each ventilation system must exhaust some air.
As part of the discussion include the design features which will permit rerouting of all exhaust air flow to remove the racioactive particles prior to exhausting into the exhaust plenum.
The staff requires the main steam and main feedwater isolation valves 410.82 to fail-closed, not fail-as-is as indicated on FSAR Figure 10.4.1.
(10.3) 00.4.7)
Verify that these valves will fail in the closed position.
RSP O
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113.e3 Table 10.1-3 referenced in Table 10.3-1 is n'ot in the FSAR.
Provide
- 10. 3:
Table 10.1-3.
213.24 Verify that the air operated modulating and on-off' dump valves and the
3 rain steam isolation valves are supplied by the Nuclear Instrur.entation Air Syste.T.
413.85 Table 10.1-1 in the FSAR specifies the stretch capacity of the steam (10. 3) generators to be 17.16 MLBS/hr.
FSAR Table 10.3-1 soecifies the
. guaranteed maximum capacity to be 15,431,353 lbs/hr.
Both of these steam rates exceed the total capacity of the main steam isolation valves, which is 16.4 MLBS/hr. as per FSAR Tab 7 ' 10.3-1.
As the result of the main steam isolation valves being a flow restriction, verify that excessive steam' flow through the valves will not result in additional wear of the valve interr.als such as to 1) prevent closure of the valve or 2) increase the~ leakage rate.
I',0. E 5 provide a detailed evaluation of the effects of a postulated failure
,1 :. '. E )
c' :ne ex;ansicn joint in tne circulating water system assuming a couble-ended guillotine break.
Two specific cases should be addressed
- 1) the effects when the designed automatic and/or operator actions are taken and 2) the effects when no actions are taken and the pumps continue
.=c operate until the water level reaches equilibrium in all affected areas.
The evaluation should include a discussion of the following considerations:
1.
The capability of detecting a failure and the means of uniquely distinguishing this type of failure to the operator.
2.
Provide a rate of rise and tctal height of the water for ca:'
affected area until equilibrium is reached.
3.
Fo each potentially flooded space, provide a discussion and drawing of the protective barrier provided for all safety-related systems that could be affected in the event of flooding.
Include in your discussion the consideration given to passagevays, pipe chases, and/or the cableways connecting the flooded spaces to the spaces containing safety-related systems or 'ccaponents outside the turbine builcing.
Discuss the effect of the ficod water cm all poter.tially submerged safety-related electrical systems and components.
4.
No credit shall be taken for isolation valve closure unless these valves are designed to safety grade requirements.
t 5.
No credi.t shall be taken for any doors or barriers which are not watertight, including wail penetrations for electrical conduit or piping.
1
%.. 5.
If credit for protection of other areas 'is taken by vir ue of the distance from the turbine building, the topographic layout should be provided which clearly defines the path of the water leaving the turbine building and the limits of water path.
Previde a complete FSAR section of the Plant Makebp Water System with 31 0.57
'i:1.5) ali su se::icns as provided for the cther systems, such as the safety evaluation section.
410.88 State how Branch Technical Position ASB 10-2, " Design Guidance.for (10. 4. 7 )
Water Hammers in Steam Generators with Top Feedring Designs" is met.
~
Discuss the design features to minimize water hammer and the con-firmatory tests to be performed.
While the Branch Technical Position addresses a specific type of steam generator, our concern,is that following any design basis event, the required auxiliary feedwater system flow might result in damage, due~ to water hammer, to the auxiliary or main feedwater system as well as to the steam. generator.
410.E?
Provide a discussion and drawings of the hydraulic control systen(s)
(10. a. 7 )
for the main steam and main feedwater isolation valves and the (i0.3, auxiliary feedwater steam inlet valves.
Include in the discussion (10. 4. 9 )
the type of fluid, the systems seismic classification, quality group class, the motive power source for the hydraulic fluid along with its
,,, seismic qualification, quality group class and whether the power source is AC or DC and its ~IEEEjClass..
~410.90 Verify tr.at the electricai power for the electro-hydraulic controllers (10.1. 7) for the main feedwater isolation valves is seismic Category I,'IEEE Class lE and th'at both valves on the same feedwater line'are not fed from the same lE source.
410.91 Provide a discussion of the design provisions in the condensate.and (10.4.7) feedwater system which provide the capability to detect and control
~
leakage frca the system.
a10. 92 Provide a response to our April 24, 1980 generic letter concerning the (10.a.9) auxiliary feedwater system.
410.93 The FSAR does not specify the cmount of water recuired for the "O.a.9) auxiliary feedwater, system.
It is the staff's position that the RSP auxiliary feecwater system tust have sufficient water capacity for four (4) hours of cperation at hot stancby prior to initiation of cooldown, as indicated in Branch Technical Position RSS 5-1.
Modify the FSAR to include the required quantity of water for the four hours at hot standby and the cooling down period until initiatlon of the decay heat removal system.
410.94 Discuss whether the auxiliary feedwater pump capacities listed in (10. 4. 9)
Table 10.4-20 of the FSAR (600gpmforeachmotordrivenand1,325gpm for the turbine driven pumps) include allowance for wear.
O
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O %
u '
-41:.95 Describe the design provisions which pronibit the reserved quantity
- 10.c.9) of auxiliary feedwater (330,000 gallons) from being used by any other system.
Does this reserve volume include the unusable. water volume in the bottom of the tank?
21 :.95 The FSAR indicates an alternate water source for the auxiliary feed-
', :. 4. 9 )
water system is the nuclear service water tank.
Provide a revised FSAR Figure 10.4-6 which shows the connection between the auxiliary feedwater system and the nuclear service water system.
410.97 In the FSAR, repeated use of the phrase " hot standby for an extended (10.4. 9) period" is made.
Define this phrase in hours.
State the dedicated water v'lume for the auxiliary feedwater system in 410.93 o
(10. 4. 9 )
the condensate storage tank and in the nuclear service water tank.
Provide the design provisions for preventing this dedicated water volume from being used by other systems.
410.99 The FSAR is unclear' with resp'ect to connecting a motor driven auxiliary (1:. 4. 9) feedsater pump to feed the steam generator to which it is not normally aligned.
FSAR Figure 9.4-6 indicates the alternate flow path can be opened by an air operated remote manual valve.
Hcwever the normal flow path cannot be isolated except by local manual operation after wounlocking the valve.
Verify whether this is correct.
4 :.100 Discuss hcw a low water ievel in the demineralized water tank indicates (10.4.9) excessive system leakage when the auxiliary feedwater system is being used.
Include how the tank level identifies the location of the leakage, as indicated in the FSAR.
a i O. i :'.
Tne FSAR states, with respect to cne auxiliary feedwater pumps, the (10. 4. 9 )
failure of one subsystem is not expected to affect the operation of the other subsystem.
Verify that the phrase "not expected to affect" means that 4t will not affect the other subsystem.
If this is not the meaning, provide a discussion of each failure which would or r.ight be expacted to affect ancther subsystem cf the auxiliary 'eed-l water system.
4'0.102 Provide a discussion of the design provisions that permit inservice
- .i.9) in;pecticr. and testi.g cf the sys:cr ccmpor.ents and a description of the inservice inspection and testing program.
t