ML19206B014
| ML19206B014 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 03/18/1974 |
| From: | Anthony Giambusso US ATOMIC ENERGY COMMISSION (AEC) |
| To: | Arnold R METROPOLITAN EDISON CO. |
| References | |
| NUDOCS 7904210714 | |
| Download: ML19206B014 (56) | |
Text
.
lhh Dockct.;e. 50-329 /'
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b-Metropolitan Edison Company AT.i :.'!r.
R. C. Arnold Vice President P. O. Box 542 Reading, Pennsylvania 19503 Ccntle=en:
On ?6a ruary 15, 1974, we received your tendered application for a licanse to operate the Three Mile Island ::uclear Station,
'Cnit 2.
In accordance with the Coa =1ssien's reculations, Sec-tien 2.101,10 C?n Part 2, we have conducted an acccatance review of your apolicatico.
On the basis of this reviev, we have concluded diat the Final Safety Analysis Report (?SAR) and the Financial Infer.a-tion are suf ficiently completa to permit us to initiate our review in most, but not all, sections.
Accordingly, you should file the appropriate nedber of copies of the applicarica as required by Section 30.30(c) of 13 CFR Part 30 as soon as nessible and subtit the 2dlitional informatice as requested in this letter.
Ouring our acceptance raview of your FSAR, wa identified addi-tional information th at will be required to conduct ene safety ravira.
The inforsation being requested is lis ted in the enclo-sur s.
To per-it review scheduling and to avoid delay tn our rerira of th2 FSA2, the infor=ation in Enclosure 1 aust be suc-
=it Ad with tn thirry (30) days folle41ng the docketing.
If yea canot sub=it this information within that time, please inf o rm us, at die time of docketing, of the schedule you intend to reet so that we may schedule appropriate portiens of the s afety revies accordingly.
The infor=ation requestad in. While resulting from tha acceptance review, is in addition to the infornatica requirements of the Standard Format and Centent Guide, n tober 19 72.
To avoid delay in our e
containment sys te:s reviev, this informatica cust be submitted within thirty (30) dafs following the docketing.
If you cannot su5=it this informatica within that t1:e, please inform us, at the ci=e of dociating, of the schedule you intend to meet,
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,N 790421 0 7/
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'!2trcyclitan Edia ca Co caw e.new, arended, or supple ental Envirennental 7ancrt f or the
trae !!ile Island '.:uclear Statica, Unit 2, cill be ccuired to update the Tr.vironmental 2cport for Units 1 and 2 filed o:
Decarber 20, 19 71, as 2nended.
Taese additienal informatica rwuirements are given in Enclosure 3.
Based on the present estinated fuel load date, this informaticn sheuld he submitted by Ap ril 1973.
If this date is not s atis fc ory, please inforn us, at the tima of dockett q, of the schedule you intend to meet.
If, during the course of our review of the scolication for the Tarce lille Island :7ucicar Statten, Unit 2, vou believe there is a naed to bring to the attentien of the Director of Licensing
=atters which involve an appeal by you of staff decisions relating to rour acelication, you may have the opportunity by contactine tha Director, the reputy Director for Technical '.evie r, er ce.
In na%ing such an a peal, which may be done either orally or in writing, ycu should specify the matters to be discussed and indicate your reascas for disagreement with the staff revievers.
Tae natters being appealed vill ba discussed at a :ceting held by the Director of Licensing.
Ycur Ccapany should be represented by the responsible corporate of ficer.
':taff representatives will include the Deputy Directors of 22 actor Projects and Technical Enview or their Assistant Deputy Director 3.
This appeals procedure is an infort:al one, designed to allow oppor-tunity for applicants to discuss, with Licensin2 nanacement, areas of disagreement in the case revie.r.
Since rely,
C g.. : E: :i:
E's 3 3. 3,.';4 A. Gianbusso, "eruty Director for ?cactor 1-rojects Directorate of Licensing
Enclosures:
is statad cca :
Listed en page 3 SEE PREVICUS YELLCU FOR ADDITIONAL CONCURRENCES LWR 2-L.\\...
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f:tropolitan Edinon Concan? ccc:
~?U Servica Corporation "ichard *..'. Hevard, Proj ect Mana;ar 250 Cherry :i111 " cad Pars ipp any,.;ew Jersey 07054 G"U Serrice Corporatten Sonaa '.t.
Crd-4 na, J r.
Safety and Licen;ing "anager 260 Cwrr/ Hill Road Pa rsiopany,.iew Je racy 070 %
Pennsylvan:s Electric Ccnpany Vice President, Technical 1101 Borad 3 tract Johnstevn, ?cnnsylvania 15 37 Coorge Tro--b ridge, 31uire S'.t av, Pittman, Forts & Trowbrid-e 910 17th S treet, 74
'Jashington, D. C.
20006 DISTRIBUTICS:
RKlecker "
Local PDR DEisenhut e Docket file AGi a=b us s o-G " 2 : l i...,
FStMary LWR 2-2 Reading -
WRegan V AVo o re '
Jdend rief JPan::arella '
Attorney, CGC '
JRutberg, OGC '
AKenneke -
ABrait=an, OAI R0 (3) -
NDub e -
3Washburn /
MSe rvice CMiles /
TR Assis tant Directors -
TR 3 ranch Chiefs -
LWR 1 and 2 3 ranch Chiefs -
ACRS (16) -
DMuller -
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Farm d C..eid i R ev. M 3, A E Cat
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"atronolitan Edisen Cc:pany.i ne r, amended, or supple: ncal Znvironnea.tal yf$ ort for the
- hree ::ile Island ::uclear S ation, Uni 2, v1V1 he reautred to update the.nvirennental Rep < -t for Units 1 cnd 2 fil2d en Decenbe r 10, 1971, as acended.
Tacsc additi/nal information requiraments are piven in Enci sure 3.
3asWd en the present estimated fuel lead data, this fpformatien/should be submitted.
by April 19 75.
If :his date is hot satisfactory, ple.ise infers us, at the tiac of dec'<eting, of 'the sch/dula you intend to coat.
\\
/
All FIEs at the theratin;; License s.yage ust be involved in develepine or installing a loose par,tv =enitor systec.
Eicher a des crip tic 1 of this system should t\\e included in the TSAX or you should provida, at the time ofjdockating, a schedul-for the revision of the FSAR to include,Jth.is information.
i
/
if, durin;; the course of our revier of the accliention for the
':hree Mile Island Nuclear Statien ' Uni 2, you believe there is a need to bring to the attention of the{ Director of Licensing satters which involve an appeal y ycu if staff decis tons relatin-te your application, you cay haye the ofpertunity by contac:ing the Director, the Deputy Director for Technical Reviev, or ne.
In taking such an appeal, whicif =ay be dbne either orally or in vritin;, ycu should specify thq matters :o be discussed and indicata your reasons for disagree =ent v.th the staff reviewers.
Tae matters beine a pealed will. be discursed at a meetin; held by the Director of Licensing. Tour Coachnv should be represented by the responsible corporate o ficer.
E-aff representatives vill include the Deouty Direct rs of Re ctor Projects and Technical naview or their Assistant Depus - Dire ors.
"*:is anneals pro-cedure is an infor.al ene, desis ed _o alle-> ooportunity for applicants to discuss, with Licensing management, arans of dis-a:;raccent in the case revicv.
Sincerely, A. Oia=b usso, Caputy Director for Reactor Projects Directorate of Licensing Enclosurns:
As stated
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UNITED STATES
- N7 ATOMIC ENERGY COMMISSION l *i WASHINGTON, D.C.
20545
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March 18,1974 e
Docket No. 50-320 Metropolitan Edisca Cc=pany ATIN:
Mr. R. C. Arnoic Vice President P. O. Box 542 Reading, Pennsylvania 19603 Gentle =en:
On February 15, 1974, we received your tendered applicatica for a license to operate the Three Mile Island Nuclear Station, Unit 2.
In acccrdance with the Cc==issien's regulations, Sec-tien 2.101, 10 CFR Part 2, we have conducted an acceptance review of your applicatien.
On the basis of this review, we have concluded that the Final Safety Analysis Report (FSAR) and the Financial Infor=a-tion are sufficiently co=plete to permit us to initiate our review in most, but not all, sections.
Accordingly, you should file the appropriate nu=ber of copies of the application as required by Section 50.30(c) of 10 CFR Part 50 as soon as possible and aubmit the additional information as requested in this letter.
During our acceptance review of your FSAR, we identified addi-tional infor=ation that will be required to conduct the safety review.
The information being requested is listed in the enclo-s ures.
To permit review scheduling and to avoid delay in our review of the FSAR, the information in Enclosure 1 must be sub-
=itted within thirty (30) days following the docketing.
If you cannot sub=it this infor=ation within that ti=e, please infor= us, at the time of docketing, of the schedule you hi%nd to meet so that we may schedule appropriate portions of inc safety review accordingly.
The information requested in, while resulting from the acceptance review, is in addition to the infor=ation requirements of the Standard Format and content Guide, October 1972. To avoid delay in our centain=ent systems review, this information must be submitted within thirty (30) days following the docketing.
If you cannot submit this informatien within that ti=e, please inform us, at the time of docketing, of the schedule you intend to meet.
M'~C3
G Metropolitan Edisen Co=pany A new, amended, or supplemental Enviren= ental Report for the Three Mile Island Nuclear S tation, Unit 2, will be required to update the Environmental Report for Units 1 and 2 filed on December 10, 1971, as amended.
These additional infor=ation require =ents are given in Enclosure 3.
Based en the present esti=ated fuel load date, this infor=ation should be submitted by Ap ril 19 75.
If this date is not satisf actory, please inform the time of docketing, of the schedule you intend to meet.
us, at If, during the course of our review of the application for the Three Mile Island Nuclear Station, Unit 2, you believe there is a need to bring to the attentien of the Director of Licensing
=stters which involve an appeal by you of staf f decisions relating to your application, you may have tha opportunity by contacting the Director, the Deputy Director for Technical Review, or me.
In making such an appeal, which may be done either orally or in writing, you should specify the matters to be discussed and indicate your reascra for disagreement with the staff reviewers.
The satters being ap. ealed will be discussed at a meeting held by the Director of Licensing.
Your Company should be represented by the responsible corporate of ficer.
Staff representatives will include the Deputy Directors of Reactor Projects and Technical Review or their Assistant Deputy Directors.
This appeals procedure is an informal one, designed to allow oppor-tunity for applicants to discuss, with Licensing =anagement, areas of disagreement in the case :eview.
Sin ce re ly,
. Giambusso, Dep"
- wr for Reactor Projects Directorate of Licensing
Enclosures:
As stated ces :
Listed on page 3 G4'C23
Metropol2. tan Edison Company ces:
GPU Service Corporation Richard W. Heward, Project Manager 260 Cherry Hill Road Parsippany, New Jersey 07054 GPU Service Corporation Thomas M. Cri= mins, Jr.
Safety and Licensing Manager 260 Cherry Hill Road Parsippany, New Jersey 07054 Pennsylvania Electric Cc=pany Vice President, Technical 1001 Borad Street Johnstown, Pennsylvania 15907 George Trewb ridge, Esquire Shaw, Pittman, Potts & Trowbridge 910 17th S t ree t, NW Washington, D. C.
20006 h'l' C ))
ENCLOSURE 1 ADDITIONAL INFORMATION REQUIREMENTS THREE MILE ISLAND NUCLEAR STATION, UNIT 2 FINAL SAFETY ANALYSIS REFORT G4-C'!5
qgh ADDITIONAL INFORMATION REQUIREMENTS THREE MILE ISLAND NUCLEAR STATION, UNIT 2 DOC KET NO.90-320 QUESTION AND COMMENT IDENTIFICATION:
1.
Numerals preceding the virgule (/) identify the AEC Licensing Branch concerned with the question, comment or response and identify the question, etc. by number.
(Branch identification precedes the hyphen (-) and the question identification follows.)
2 Numerals following the virgule identify the FSAR section being referenced.
BRANCH IDENTIFICATION:
1 Projects 2
Mechanical Engineering 3
Materials Engineering 4
Structural Engineering 5
Reactor Systems 6
Electrical Instrumentation and Control Systems 7
operstional Safety 8
Auxiliary and Power Conversion Syctems 9
Containment Systems 10 Quality Assurance 11 Effluent Treatment Systems 12 Accident Analysis 13 Site Analysis 14 Radiological Assessment 15 Cost Benefit Analysis 16 Core Performance b' d.' C 2 {2'
t 1-2
~
ADDITIONAL INFORMATION REQUTREMENTS THREE MILE ISLAND NUCLEAR STA? ION, UNIT 2 DOCKET NO.60-120 1-1 / 1.2
" ould include brief discussion of the principal design criteria, operating characteristics and safety considerations for the radioactive waste management system.
Should have a secticn in which systems and equipment being shared with Unit 1 are identified.
1-2 /1.2.1 Site ownership information should be included.
Control of access to the site should be discussed briefly.
1-3 /1.3.1 Table 1 3-1 should include a comparison of radioactive waste management systems.
1 4 /1.3.2 Missing cross references to the appropriate saction in the FSAR that describes the changea and tha reacone
'n" *ba- *beuld ha enrrliei.
1-5 /1.5.7 last 'i e -
"arial shipping red 'nsamblie "
1-6 /1.6 This tabulation should include the applicable sectians of the PSAR in which the tabulated reports are refer-enced.
NOTE:
Some of the B&W Topical Reports listed have not been approved.
If these are n support of this application and they are determined to be deficient upon review by Licensing, acceptable responses will have to be submitted by the applicant or 3&W before review of this SAR can be completed.
1-7 /1.7 Compilation of the abbreviations and terms used in the SAR text should be included.
1 -8 /3.1 Two sets of criteria, Appendix A, 10CFR50, July 11, 1967, and Appendix A, 10CFR50, July 15, 1971, are included.
Discussion texts of the two sets are identical in some cases and differ in others.
The facts stated in these discussions should be brought into agreement.
1-9 /3.1.2 This section states that the discussion text includes an explanation of significant differences between the actual design and the AEC General Criteria dated July 15, 1971.
The text should identify and discuss these differences, if they exist.
1-10/3.6 NOTE:
Appendix 3F is not included and is promised by 5/1/74 in the FSAR transmittal letter.
OF C37
1-11/
NOTE:
Amendment promised the first quarter of 1076 3.7.3.14 1-12/
N0!Ze A " p e r.
.a 3: was not included.
- 3. A.1. 6. 9 1-13/3.11 NOTE:
A-adment promised the first ouarter of 1974, 1-14/9.3.3 NOTE:
failed fuel detection system--Amendment te be flied no later than the end of 1974 1-15/
NOTE:
Loss of inst ument air--Information tn be 15.1.32 filed in the second quarter of 1974 1-16/16.3.3 NOTE:
Technical Specification on the service water system should be included.
1-17/16.
NOTE:
Technical Specification on containment integrity should be included.
1-18/
NOTE:
Technical Specifications should include:
16.3.5.2 Shutdown reactivity Rod insertion limits Inoperable rod position indicator channels Rod drop time Rod position deviation monitor 1-19/
NOTE:
Technical Specifications on Surveillance of 16.4.4.4 the Hydrogen Recombiner are to be supplied during the third quarter of 1974 E4-03S
~
ADDITIONAL INFORMATION REQUTHEMENTS THREE MILE ISIAND NUCLEAR STATION, UNIT 2 DOCKET NO,90-120 5.
REACTOR COOLANT SYSTEM Inservice Insnection Procram for ASME Code Class 2&3 Comnonents 3-1/5.2.8 Provide sufficient information about your proposed TS 4.2 ESP inservice inspection program to indicate that the program will provide a degree of assurance of system integrity comparable to the program recom-mended in Regulatory Guide 1.51, " Inservice In-spection of ASME Code Class 2 and 3 Nuclear Power Plant Components," May 1973.
Inservi e Testine of Punus and Valves 3-2/5.2.8 TS 4.2 Provide your program for the inservice testing of pumps and valves that are part of ASME Boiler and Pressure Vessel Code,Section III, Class 1, 2, and 3 safety related systems.
Furnish sufficient infor-mation to indicate that the pump test program fol-lows the provisions of ASME Section XI, Summer 1973 Addenda, "ubsecticn IWP, and that the valve test program follows Subsection IWV, as far as practical.
3-3/5.2.8 Identify all pumps and valves in the program.
The TS 4.2 component identification may be either in tabular form of reference P&I drawings, or a combination of each method.
Each valve identity should also include the I#V-2101 Category designation.
Flywheel Intecrity 3 4/5.2.6 Provide sufficient information about the primary coolant pump-motor flywheels to indicate the degree of flywheel integrity comparable to the program recommended in Regulatory Guide 1.14,
" Reactor Coolant Pump Flywheel Integrity",
October 27, 1971.
E4-C'!9
1-5 ADDITIONAL INFORMATION R2QUIREMENTS THREE MILE ISLAND NUCLEAR STATION, UNIT 2 DOCKET NO. 50-120 4-1/3.7.4 Seismic Instrumentation It is a Regulatory Staff position that in the seismic instrumentation program, one multielement seismoscope should be installed at the basement of the containment for a rapid determination of the ground input response spectra during an earth-quake.
You should revise your program accordingly.
4-2/3.8 Structural Desien Discuss the degree of conformance with the pertinent Regulatory Safety Guides:
1.13 Fuel Storage Facility Design Basis 1.16 Structural Acceptance Test for Concrete Primary Reactor Containments 1.19 Nondestructive Examination of Primary Containment Liner # elds 1.46 Protection Againct Pipe Whip Inside Containment 1.55 Concrete Placement in Category I Struc-tures You t.ould list and discuss the deviations, if any, from the Regulatory Guides above.
4-3/3.8 If local building codes are utilized, provide sufficient supporting information to allow an evaluation as to the applicability of the document and its degree of conservatism.
4 h/3.8 The structural criteria for evaluating the design of Seismic Category I etructures inside or out-side of containment are not complete.
An accept-able set of criteria is contained in the nttached Document (3).
Sufficient information should be provided to establish the extent of compliance with these design criteria.
Where inconsistencies or deviations from these criteria are proposed, justification should be provided to demonstrate that your criteria are equivalent with respect to the applicable safety margins.
E4-040
1-6 DOCUMENT (B)
STRUCTURAL ENGINEERING BRANCH DIRECTORATE OF LICENSING STRUCTURAL DESIGN CRITERla FOR EVALUATING THE Ettt. CTS OF HIGH-E'IERGY PIPE BREAYS CN CATECORY I STRUCTURES OUTSIDE THE CONTAIMIE:.T CONTENTS A.
Introduction B.
Loads, definition of terms and nomenclature C.
Acceptab.1.e load combinations and allowable limits for Category I concrete structures D.
Acceptable load combinations and allevable limits for Category I steel structures E.
Acceptable procedures for determination of the effect of an i=pacting whipping pipe on concrete and steel structures F.
Acceptable procedures for design of structural pipe restraints June 1973 G4- -041
1-7
- A.
INTRODUCTION General Design Criterion 4 of Appendix A to 10 CFR Part 50, "Jeneral Design Criteria for Nuclear Power Plants," necessitates that structures important to safety, classified as Category I structures, shall be designed to accoc=odate the ef fects of, and to be compatible with, the environmental c onditions associated with normal operation, =aintenance, testing and postulated accidents.
These structures shall be appro-priately protected against dynamic effects, including the effects of missiles, pipe whipping, and discharging fluids associated with postu-laced high-energy pipe rupture accidents and from events and conditions outside the nuclear power unit.
This document presents a set of acceptable criteria for evaluating and assuring the required protection.
It is assu=ed that the follow-ing steps, which are not structural in nature at:d are thus not within the scope of this document, have already been p2rformed and the neces-sary design parameters already defined:
1)
Systems in which pipe breaks are postulated and for which pro-tection against the effects of such breaks should be provided, have been defined, 2)
Locations of postulated breaks and type and orientation of each break, guillotine or longitudinal, have ~oeen determined,
.G4-04L
1-8 3)
Protection criteria for each postulated break have been estab-11shed.
This should identify the structures, systems and ccc-ponents to be protected fro = the effects of the break, and 4)
All induced loadings for each postulated break are defined, including:
a) Differential pressure across compart=ents, if any, as a function of ti=e, b)
Jet i=pinge=ent force, if any, on a protective barrier, as a function of ti=e, and c)
Whipping pipe i= pact p'.ra=eters, if any, on a protectiva barrier or a pipe restraint, including the equivalent = ass, i= pact area and i= pact velocity.
B.
LOADS, DEFINITION OF TERMS AND NCMENCLATL*RE The following no=enclature and definition of ter=s will apply :: all the criteria that follow in this docu=ent.
A'.1 the =ajor loads to be encountered and/or to be postulated during a high-energy pipe rupture event are ' listed.
All the loads listed, however, are not necessarily ap,11 cable to all the structures and their ele =ents in a plant.
Loads and the applicable load cc=binations for which each structure has to be checked and evaluated will depend on the conditions to which that particular structure could be subjected.
64--043
1-9 B.1 NORMAL LOADS Normal lo' ads are those loads to be encountered during nornal plant operation. They include the following:
D ---- Dead loads and their related noments and f(rces, including any permanent equipment loads, and prestressing loads, if anv.
L ---- Live loads, present during the pipe rupture event, and their relat.ed moments and forces.
T ---- Thermal loads during normal operating conditions.
R Pipe reactions during normal operating conditiens.
B.2 SEVERE DNIR02 ENTAL LUADS Severe environmental loads are those loads that could infrequently be encountered during the plant life.
Included in thic category are:
Fego - Loads generated by the Operating Basis Earthquake or, if an OBE is not specified, loads generated by half the Safe Shutdown Earthquake.
If both are specified, they shall be the largest of the two.
B.3 EXTREME ENIRO:CfENTAL LOADS Extreme environmental loads are those loads which are credible but are highly improbable.
They include:
Fegs - Loads generated by the Safe Shutdown Earthquake.
- 1
1-10 B.4 ABNORMAL LOADS Abnor:21 loads are those loads generated by a postulated high-energy pipe break accident within a building and/or ecmpartment t.her eo f.
Included in this category are the following:
?, -- Pressure equivalent static load within or across a ce=part-ment and/or building, generated by a postulated break, and including an appropriate dynamic factor to account for the dynamic nature of the load.
T, --- Thermal loads under thermal conditions generated by a postulated break and including T.
R -- Pipe reactions under ther:1al conditions generated by a postulated break and including R.
Y
- Equivalent static load on a structure genereted by the reaction on the broken high-energy pipe during a postulated break, and including an appropriate dynamic factor to acccun:
for the dynamic nature of the load.
Jet impingement equivalent static load on a structure gen-Y) erated by a postulated break, and including an appropriate dynamic factor to account for the dynamic nature of the load.
Y Missile impact equivalent static load on a structure gen-ersted by or during a postulated break, like pipe whipping, and including an appropriate dynamic factor to account for the dynamic nature of the load.
GVC?N
1-11 6-In determining an appropriate equivalent static load for P, Y, Y and Y, elasto-plastic behavior =ay be assumed with appropriate ductility ratios and as long as excessive deflections will not result in loss of function.
B.5 OTHER DEFINITIONS S
For structural steel, S is the required section strength based on the elastic design methods and.the allowable stresses defined in Part 1 of the AISC " Specification for the Design, Fabrication and Erection of Structural Steel for Buildings,"
February 12, 1969.
U ---- For concrete structures, U is the section strength required to resist design loads and based on methods described in ACI 318-71.
Y
- For structural steel, Y is the section strength required to resist design loads and based on plastic design methods described in Part 2 of AISC " Specification for the Design, Fabrication and Erection of Structural Steel for Buildings," February 12, 1969.
C.
LOAD COMBINATIONS AND ACCEPTANCE CRITERIA FOR CATEGORY I CONCRETE STRUCTURES The following presents an acceptable set of load c.ombinations and allowable limits to be used in evaluating and checking Category I concrete structures outside the containment for the effects of e
1-12 high-energy pipe breaks.
Concrete barriers, used to provide a shield against the ef fects of high-energy pipe breaks, will have to maintain their structural integrity under all credible loading conditions. To assure that the structural integrity will be maintained, limits on the required strength capacities are recommended.
C.1 LOAD COMBINATIONS The following load comi". nations should be satisfied:
1)
U=D+L+T
+R
+ 1.5 P a
a a
2)
U=D+L+T
+R
+ 1.25 P + 1.0 (Y
+Y
+ Y ) + 1.25 Fego a
a a
r a
3)
U=D+L+T
+R
+ 1.0 P + 1.0 (Y
+Y
+ Y ) + 1.0 Feqs a
a a
r m
The maximum values of P, T, R, Y, Y and Y, including an appro-a a
a j
r a
priate dynamic factor, shall be used unless a time-history analysis is performed to justify otherwise.
Both cases of L having its full value, possibly present during the pipe rupture event, or being ccmpletely absent should be checked for.
Por combinations (2) and (3), local stresses due to the concentrated loads Y ' Y and Y, may exceed the allowables provided there will be r
j no loss of functica of any adfety-related system.
Existing structures will have to be checked and evaluated for the above three ccabinations.
The failure capacity of concrete structures e
1-13 may be checked by using the " Yield Line Theory." The co=bined loads should not exceed 90% of the calculated failure capacity.
In such situations, however, it should be verified that neither excessive deflections nor excessive cracking, will result in the loss of func-tion of any safety-related system.
D.
LOAD COltB1 NATIONS AND ACCEPTANCE CRITERIA FOR CATECORY I STEEL STRUCTURES Category I steel structures outside the contain=ent, whose function is to provide protection against the effects of high-energy pipe breaks, will have to maintain their structural integrity under all credible loading conditions. To assure this, limits on resulting stresses or required strength capacities are recom: ended.
D.1 LOAD CC!!31 NATIONS a)
If elastic working stress design methcds are used:
1) 1.6S=D+L+T
+R
+P a
a a
2) 1.6S=D+L+T
+R
+P
+ 1.0 (Yj+Y + Y ) + Feq:
a a
a r
a 3) 1.6 S = D + L + T
+R
+P
+ 1. 0 (Y
+Y
+ Y ) + Fegs a
a a
j r
m b)
If plastic decign methods are used:
1)
.90 Y = D + L + T
+R
+ 1.5 P a
a a
+ Y ) - 1.25 Feq:
2)
.90 Y = D + L + T
+R
+ 1.25 P + 1. 0 (Yj+Y a
a a
r
=
3)
.90 Y = D + L + T
+R
+ 1.0 P + 1.0 (Y
+Y
+ Y ) + 1.0 Feqs a
a a
r
=
In combinations D.l(a) and (b), thermal loads can be neglected when it can be shown that they are secondary and self-limiting in nature and whera :he
=aterial is ductile.
(i4 Tag
1-14
_9-In co=binations (1), (2) and (3), the maximum values of P,, T,, R,,
and'Y including an appropriate dynamic f actor, shall be used Y), Y r
unless a time-history analysis is performed to justify otherwise.
Both cases of L having its full value, possibly present during the pipe rupture event, or being completely absent should be checked fer.
For cctbinations (2) and (3), local stresses due to the concentrated loads Y, Y) and Y may exceed the allowables provided there will bc no loss of function.
Furthermore, in cceputinr, the required sectica strength, S. the plactic section modulus of steel shapes may be used.
Existing structures will have to be checked and evaluated for the above three combinations.
The 0.90 reductien factor applied en the required section strength, Y, can be increased to 1.0.
In such situa-tions, however, it should be verified that excessive deflections will not result in the loss of function of any safety-related system.
E.
ACCEPTABLE PROCEDURES FOR DETER:tINATIOM OF THE EFFECT OF A:: I:4PACTI: G WIPPI:;G PIPE ON COMCPITE A!!D STEEL STRUCTURES If pipe whipping is permitted and if the whipping pipe can impact a
barrier whose structural integrity has to be maintained during and after the event of the pipe rupture, the barrier will have to be designed to resist that impact.
Essentially, the inpacting pipe can be considered as a missile for which the following parameters can be defined; impact velocity, impact equivalent area and equivalent e
~
1-15
- lo -
cass of the missile.
Procedures used in determining these parameters are outside the scope of this document.
Missile barriers, whether concrete or steel, should have suf ficient strength to stop the postu-lated missile. To acccmplish this objective, prediction of local and overall damage due to missile irract is necessary.
Local damage prediction, in the i= mediate vicinity of the impacted area, includes esti=ation of the depth of penetration and whether secondary missiles might be generated by spalling in case of concrete targets.
Overall damage prediction includes estimation of the struc-tural response of the target to the missile impact, including struc-tural stability and deformations.
E.1 LOCAL DAMAGE PREDICTION a)
In Concrete There are several empirical equations available to estimate misaile penetration into concrete targets. The most commonly used is the modified Petry equation, as given by A. Amirikian in
" Design of Protective Structures," Bureau of Yards and Docks, NP-3726 (1950). This equation, having been widely used, is presently acceptable.
Should other equations, however, be used, the level of conservatism in these equations should be ccmparable
'to that of the Lolified Petry equation.
Ac ual testing for deter-mining penetration in concreta is acceptable.
e E4-C50
1-16
' b)
In Steel' Extensive series of tests were conducted by the Stanford Research Institute on penetration of missiles into steel plates. The results of these tests were su==arized by W. B. Cottrell and A. W. Savolainen in Chapter 6 of Vol. 1 of U. S. Reactor Contain-ment Technology, ORNL-NSIC-5.
Equations for penetration of missiles into steel plates presented in this chapter, having been widely used, are presently acceptable.
Should other equations, however, be used, the level of conservatism in these equations shall be comparable to that of those centioned above.
Actual testing for determining penetration in steel is acceptable.
E.2 OVERALL DAMAGE PREDICTION The response of a structure to a missile impact depends largely on the location of impact, e.g., midspan of a slab or near the support, on the dynamic properties of the target and =1ssile and on the kinetic energy of the missile.
In general, it will be conservative to absorb all the missile kinetic energy Lato structural strain energy in the target.
However, energy losses due to missile deforma-tion and local penetration =ay be accounted for.
Af ter a check has been =ade on whether the missile will penetrate the barrier or not, an equivalent static load can be deter =ined from which the structural response, in conjunction with other loads that e
G4-051
1-17 can then be evaluated using conventional methods.
might be present, An acceptable procedure for such an analysis is presented in a paper by Williamson and Alvy, of Hol=es and Narver, Inc. entitled " Impact Effects of Fragments Striking Structural Elements," tiF-6515 (1957).
Should other =ethods be used, hcwever, the level of conservatisa in of those mentioned above.
these methods should be comparable to that ACCEPTABLE PROCEDURES FOR DESIGN OF STRUCTURAL PIPE RESTRA F.
Protection of Category I structures, systems and cceponents from the dynamic effects of postulated high-energy pipe ruptures can be accomplished in some situations by providing pipe restraints in These restraints should critical locations on the piping syste=s.
function =ainly by preventing the ruptured pipe, or portions thereof,
from becoming a missile that might impact and damage other critical systems, and by preventing the ruptured pipe frem whipping and impac:-
The ing critical systems not capable of resisting such an i= pact.
o.' dead and live load supports and of restraints =ay be independent However, should a pipe whip restraint be intended seismic restraints.
to function also as an operating dead load and/or seismic restraint, all applicable loads should be considered in the design of the restr: int.
F.1 ANALYSIS METHOD The structural analysis of pipe restraints may consict of an energy-balance approach, where a potential collapse mechanism is first e
9 t-19 established. The displacement of this mechanis= will reach its limit, by conservation of energy principles, when the external work available equals the internal work done on the restraint.
Cyternal work expressions may include kinetic expressions where mass and velocity of the ruptured pipe are known.
Internal work expressient are grtphically represented by the area under a resisting force-displ:Acc=ent curve.
F.2 ALLOWABLE YIELS STRENGTH Due to the high rate of strain that the structural restraint would experience after pipe rupture, and partly due to the strain-hardening effects, the static yield strength of the =aterial used =ay be increased by 15%.
F.3 ALLOWABLE STRAINS In general, strains of up to 50% of ultimata strain are acceptable, provided there is no loss of function.
Where buckling is critical in co=pression =e=bers, the load on the members should be li=ited to 90% of the buckling load.
F.4 GAP EFFECT Where gaps are provided between pipes and restraints, the kinetic energy of the pipe i=pacting the restraint =ay be critical and should not be ignored.
Moreover, the kinetic energy of the pipe after rebound may be = ors critical and should also be considered.
E4-053
1-19 14 -
F.5 ANCHOR DESIGN Pipe res'traints should be anchored in concrete and/or steel structures.
Strains and/or stresses induced in the structure by loading the restraint should be considered and et.e design of the structure should be checked in accordance with criteria already presented in this docunent.
e Y
>s
1-20 ADDITIONAL 1.
ORMATION REQUIREMENTS THREE MILE TSIAND NUCLEAR STATION, UNIT 2 DOCKET NO, <0-120
<-1/15,0 Table 15-1.
The FSAR should ba reviced tc imindo analyses of a spectrum of postulated feedwater syntem piping breaks, 4.ncide n.nd outside contain-m.ents as part of event 14..
G4-055
JITICNAL INFORMATION REQU. _AENTS 1-21 THREE MILE ISLAND NUCLEAR ST ATION, UNIT 2 DOCKET NO.40-320 6-1/1.1 The physical plan layout for the upper floors of the Control Service Building is not provided in the FSAR (See Figure 1.2-8) and is required for our review.
In addition, both Firures 3.2-18 and 1.2-19 show the same elevation section for this building but have slightly different features.
Provide the correct figures for this building.
6-2/3.11 Section 3.11 states that "this section will be completed and incorporated as an amendment during the first quarter of 1974".
The information in this section is required for our review and its omission will not allow our review to be completed.
This material should be furnished in accordance with the promised schedule.
6-3/7.1 It appears from our examination of the ?SAR dia-crams that wiring and cable interconnection draw-ings will be required to assist in our review and in our independent evaluation of your compli-ance with the separation criteria.
We must be abl e to verify your implementation of the physical separation criteria for protective systems, com-ponents and penetrations in the plant.
The
" functional" drawings provided in the FSAR are inadequate for this purpose, s.)
Provide the typical w ming and cable interconnection drawings to permit the above evaluation, b.)
Provide a tahu.stion showing cross-refer-encing between the electrical instrumen-tation and control figure numbers and tne referenced drawings similar to that given in Section 1.7 for tite P&I drawings.
c.)
Provide specific references in the text or a tabulation by system of the individ-ual drawings in Sections 7A and 73 E4-056
1-22 ADDITIONAL INFORMATION REQUIREMENTS THREE MILE ISLAND NUCLEAR STATION, UNIT 2 DOCKET NO,60-120 7-1/13.1.3.
NOTE:
Resume for Assistant Superintendent is 2.2 indicated to be supplied at a later date.
Should provide the date by which this information will be provided.
7-2/13.1.3.
NOTE:
Resume for Nuclear Engineer is indicated 2.6 to be supplied at a later date.
Should provide the date by which this information will be supplied.
7-3/13.3.1 NOTE:
Written agreements with offsite agencies for emergency services are indicated to be supplied at a later date.
Should provide the date by which this information will be provided.
7 4/14.1.3.2 NOTE:
A majority of the test abstracts for the preoperational test phase of the startup test program are not included.
Should provide the date by which this information will be provided.
(iWTIi7
1-23 Acceptance Review for Completeness Three Mile Island Unit 2 (OL)
General Note Sections 3.5, 9.0, and 10.0 were at leas t 907. cemplete in accordance with the Standard Format.
Schematic diagrams of complex systems were legible, but only under a magnifying glass.
They could be further folded out and/or placed on numerous pages, with lines in different colors according to their classification.
Section 9.4.15, Reactor Building (Heating, Ventilation, and Air Conditioning Systems) should be in Chapter 6.2 for stricter adherence to the Standard Format.
The following comments respond to the section and title found in the FSAR:
8-1/ 9.1.1 New Fuel Storage 1.
Diagrams are requirpd,to show the new fuel storage racks.
2.
The narrative should describe the new fuel storage rack construc-tion method, and provide the results 02 an evaluation of the racks to withstand the maximum uplif t forces expected.
8-2/ 9.1.2 Spent el Storage 1.
The narrative should include a safety evaluation using the positions in Regulatory Guide 1.13, Fuel Storage Facility Design Basis, as a guide, and present evidence of equivalent safety where departure from the guide exists,.
64.~C[IS
1-24 FuelCooking and Cleanuo System 8 - 3/ 9.13 Spent 1.
The narrative should include a safety evaluation using the positions in Regulatory Guide 1.13 and 1.29 as a guide, and present evidence of equi valent safety where departures frcm the guide exist.
2.
The design basis for the demineralizer flow rate and resin deple-tion estimates, as a tunction of buildup of fission and corrosion
~
products, shcula be provided.
8 4 6.2.4 Potable and Sanitary Water Systems
/9 1.
The potable water system narrative should include a safety evalu-ation of possible radiological contamination due to failures in the piping system in addition to the possible contamination to the source as found in the existing narrative.
8-5/9. 3. 3 Eauie=ent and Floor Drainage System 1.
Provide the results of an evaluation of drainage piping elevation r.
and arrange =ent, including diagrams, to show that drainage f rom one compart=ent will not back-up Lnto a compartment containing Seismic Category I equipment.
8-69.4.1 Control Rocm (Heating, Ventilating and Air Conditiening Sys tem)
/
1.
Provi*de the design assemptions and design provisions which d ter-S mine and enable the control room recirculation system to maintain a habitable space for 40 man days.
Define this in calendar days.
{54.'~C[I3
1-25 8-7/ 9.5.1 Fire Protection System 1.
Demonstrate how equivalent safety is achieved by the present system in order to meet the diesel oil storage requirements of IEEE Std. 308-1971.
8-8/ 10.3 Main Stean Suoply System 1.
Provide the documentation showing compliance with AEC General Design Criterion #4, " General Information Required for Consideration of the Effects of a Piping System Break Outside Containnent",
a letter from A. Giambusso to all applicants dated December 12, 1972, and where practicable, " Criteria for Determination of Postulated Break and Leakage Locations in High and Moderate Fluid Piping Systems outside of Containment Structures", Appendix A to a letter from J. F. O'Leagrto W. G. Kuhns, President, General Public Utilities Corp., dated July 12, 1973.
(Section 3.6 reports that high energy pipe rupture data is provided in Appendix F.
Appendix F was not included in the submittal).
E4--GEO
AuDITIONAL INFORMATION REQUInaMENTS 1-26 THREE MILE ISLAND NUCLEAR STATION, UNIT 2 DOCKET NO. 4C-320 10-1/17.2 The description of Metropolitan Edison (Met Ed)
QA Program should be amended to include, in accord-ance with Part 50 34 (b) (ii), a more complete dis-cussion of how the 18 QA criteria of 10 CFR Part 50 Appendix B will be implemented throughout the oper-ations phase.
10-2/17.2 Met Ed should commit to comply with the guidance 17.3 contained in AEC's " Orange Book" (dated October 26, 1973), including ANSI Standards therein, or identify any exceptions and describe acceptable alternatives.
10-3/17.2 Met Ed should identify a position within its organ-ization with the assigned responsibility for the review of and concurrence with QA Program (s) devel-oped and/or implemented for them by other organiza-tions, including GPU Service Corporation (GPUSC).
10 hA7.2.18 Al though Met Ed has penvi dad da-cripti cr c' i t-n sudit program, tha 4a cripti nr ebnuld 4"clude pro-visions for the ennduct of periodic audits, by Me+
Ed, of the implementation of the QA Program activ-ities delegated to GPUSC and other major organiza-tional contractors.
10-5/17.2.18 The scope and frequency of the Met Ed and GPUSC internal and external audit programs should be prov$ded.
10-6/17.2.1.3 Position qualifications should be provided for Met Ed's Manager of Operational Quality Assurance.
10-7/17.2.2 The structures, systems and components covered by the QAProgram should be identified or cross-refer-enced in Chapter 17, 10-8/17.2 A listing of the titles of QA procedures applicable 17.3 to Sections 17.2 and 17 3 should be provided along with a brief abstract of their scope and purposes and a matrix or tabulation relating these to the 18 QA requirements of 10 CFR Part 50 Appendix B.
10-9/17.2 Provisions for the indoctrination and training 17.3 program for QA activities should be described.
10-10/17.2.10 Discussion of the inspection program during plant turnover and during the operations phase is inad-equate.
Criteria / bases should be included denoting when inspections are required and when they are not required.
WCO
1-27 Accident Analysis Branch Acceptance Review Ouestions Three Mile Island Unit 2 12-1/*section 1.2 Provide a plan view drawing of the control room layout similar to those shown in Figure 1.2-17.
12-2/*section 2.1 Provide a map which clearly shows the distances between release points of gaseous radioactive effluents and the site boundary.
12-3/ Section 2.2 Figure 2.1-3 indicates that the Pennsylvania railroad line is located approxinately 2000 feet (0.4 miles) east of the Three Mile Island nuclear facility.
This plot plan further shows a railroad siding entering the site via a bridge from the north east direction.
Provide the following inf or=ation with respect to rail traffic on both of these lines:
1.
Type, quantity and frequency of toxic gases that may be transported near the site.
n 2.
Provide an analysis of the effects of an accident involving any hazardous caterials regularly transported near the site on the safe operation of the nuclear facility.
Include the following:
Ed~Cf32
- Required for acepptance.
\\
1-28 (1)
Effects of delayed ignition of a cloud of propane gas on both reactor structures and ce=ponents and the operation of the diesels.
(2) release of toxic airborne chemicals on control room personnel including the ability to isolate the control room on detection of the particular substances which might be released.
(3)
Indicate the =ethod for controlling the shipment of hazardous materials on the railroad siding leading to the plant site, (e.g., using the siding as a switching facility.
12-4/*section 6. 2 Please supply the information requested in Regulatory Guide 1.70.2 relating to filtration system and su=p design.
4 12-5/ section 6.2.3 containment sorav systems A detailed description of the fission product re= oval function of the Containment Spray System should be provided in this section if the mystem is relied on to perform *his function following a design basis accideht.
6 12-6/ Section 6. 2.3.1 Design Bases This section should provide the design bases for the fission product removal function of the contain=ent spray system, including, for exa:ple:
- Required for acceptance.
- tj g
\\
9 1-29 (1)
The postulated accident conditions and the extent of si=ultaneous occurrences that deter =ine the design require =ents for fission product scrubbing of the contain=ent atmosphere; (2) A list of the fission products (including the species, of iodine) which the system is designed to remove, and the extent to which credit is taken for the cleanup function in the analyses of the radiological consequences of the accidents di; cussed in Chapter 15 of the SAR; (3)
The bases e= ployed for sizing the spray system and any components required for the execution of the atmosphere cleanup function of the system.
12-7/* f ction 6. 2. 3. 2 Svstem Design (As Affected av Fission Product Removal Function)
This section should provide a description of systems and components employed to carry out the fission product removal function of the spray system, including the method of additive injection (if any) and delivery to the containment.
Detailed information should be provided in this section concerning:
(1)
Methods ana equipment used to ensure adequate delivery and mixing of the spray additive (where applicable).
(2)
Source of water supply during all phases of spray system operation.
(3)
Spray header design, including the nunber of no :les
- Required for acceptance.
(M);-DE4
1-30 per header, no::le spacing and orientation (a plan view of the spray headers, showing nozzle location and orientation, should be included.)
(4)
Spray no: le design, including the drop size spectrum produced by the nozzle.
Source of the data, method of measurement, and expected accuracy should be discussed.
(5) A description of the operating modes of the system should be given including the tine of system initiation, time of first additive delivery through the no::les, length of injection period, time of ' initiation of recirculation (if applicable), and length of recirculatien operation.
Spray and spray additive flow rates should be supplied for each period of operation, assuming minimum spray operation coencident with maximum and minimum safety injec-ion flew rates, and vice versa.
(6)
The regions of the containment covered by the spray.
List the containment volu=es not covered by the spray, and estimate the forced or convective post-accident ventilation of these unsprayed volumes.
Indicate the extent to which credit is taken for the operability of duct work, dampers, etc.
bMOf5[5
1-31
_3_
12-8/* Section 6.2.3.3 Design Evaluation Provide an evaluation of the fission product removal function of the containment spray system. The system should be evaluated for fully effective and minimum safeguards operatien, including the condition of a single failure of any active component.
If the calculation of the spray effectiveness is performed for a single set of post-accident conditions, attention should be given to the effects of such parameters, as temperature, spray and sump pH, (and the resulting change in iodine partition),
drop size, and pressure drop across the nozzle, in order to ascertain that the evaluation has been performed for a conservative set of these parameters.
12-9/* Section 6.2.3.4 Tests and Inspections Provide a description of provisions made for testing all essential functions required for the iodine removal effectivenss of the system.
In particular, this section should contain:
(1)
A description of the tests to be perfor=ed to verify the capability of the systems, as installed, to deliver the spray solution with the required concentration of spray additives to be used for iodine re= oval.
If the test fluids are not the actual spray additives, describe the liquids of similar density and viscosity to be
- Requi:ed for acceptance.
E4-CES
1-32
-e employed, and discuss the correlation of the test data with the design requirements.
(2)
A description of the provisions made for testing the containment spray nozzles.
(3)
The provisions made for periodic testing and surveillance of the content of the spray additive tank (s).
Provide the bases for surveillance, test procedures, and test intervals deemed appropriate for the system.
12-10/
- Section 6.2.3.5 Instrumentation Requirements This section should include a der:ription of any instrumentation of the spray syste= required for actuation of the system and
=onitoring of the fission product removal function of the system.
12-11/
- Section 6.2.3.6 Materials Specify and discuss the chemical composition, concentrations in storage, susceptibility to radiolytic or pyrolytic decomposition, corrosion properties, etc., of the spray additives (if any), the spray solution, and the contain=ent sumo solutica.
6
- Required f or acceptance.
G4 CE7
\\
1-33 12-12/ *Section 6.4 control Room Habitability 1.
The flow rate of unfiltered air leaking-into the Control Room should be calculated for the condition of control room isolation.
The FSAR should include a clear description of the assumptions used in the analysis including:
a.
Identif,1 cation of leakage paths (doors ; duct, pipe, and cable penetrations ; dampers; etc.)
b.
Leakage path characteristics (pressure differential / leakage flow rate relationship)
Esti= ate of pressure differentials (caused by trind ef fects, c.
stack effects, barometric pressure variations, ventilation units servicing spaces adjacent to the control room, negative pressures at suction side of f acs). A 1/8" Hg pressure differential for all leak paths (except those exposed to the negative pressure on t!.; suction side of fans) is nor= ally considered adequate to account for these effects.
d.
Leakage contribution from all pathways (filtered and unfilteredileak rates should be reported separately).
- Required for acceptance.
(M CES
\\
1-34 2 Identification should be made of toxic materials (quantities, method of storage, etc.), such as chlorine, that may be stored on or in the vicinity of the site, which, assuming a container rupture, may interfere with control room operation.
b4~CES
1-35 12-13/*section 15 1.
An analysis of the thyroid, beta skin, and whole body L : a doses received by control room operators during accident situations should be provided.
The dose contribution from each separate source of radioactivity should be tabulated.
When evaluating the effectiveness of the control room protection features, all types of accidents should be considered; however, only the limiting accidents need be analyzed in detail.
As a minimum, calculate the doses received by the control room operators from a main steam line break accident, a loss-of-coolant accident, and a fuel handling accident.
In the case of the LOCA, allowances may be made for control room occupancy factors of 1.0 between 0-24 hours, 0.6 between 1-4 days, and 0.4 between 4-30 days.
The =ethod used to calculate the doses should be provided A complete list of assumptions and input data should be provided including:
The source terms used for eacP a.
. of release.
All potential sources of tivity including cen tainment leakage, exfiltrati-
..y, vent and stack releases,
- Required for acceptance.
E4' 070
1-36 penetration leakage and activity which may be transferred directly to the control room frcm the radwaste and turbine buildings and from other portions of the control building should be considered (See item c.)
b.
The distances between the points of radioactivity release for each design basis accident and the air intake to the control room.
c.
An evaluation of the potential for radioactive material, noxious gases, or steam to be transferred directly into the centrol room from adjacent areas and buildings.
This should include a description of all potential paths for transport such as the duct work, corridors, doorways, elevator shafts, etc.
d.
The expected dilution factors between the expected release points and the control room air intake (or other appropriate opening).
Assumptions as to wind speed and exposure frequency made during the course of the accident should be clearly stated.
Technical references and/or experimental data to justify the f actors used in the analysis should be provided.
EA f%
to 4
9 1~37 2.
It is stated on page 15.1.14-2 that an analysis of potential leakage from the engineered safety feacures during a maximum hypothetical accident is presented in Chapter 6 of the final safety analysis report.
Please indicate more precisely where this information is located as we were unable to find this analysis in Chapter 6.
The analysis should include the following:
(1)
Assume that in the event of a loss-of-coolant accident that equipment outside contain=ent is leaking at a maximum operational leakage rate (i.e., postulate a damaged seal, or packing, or some other leakage path in which leakage would be at a =axim im but not great enough to cause the pu=p or equipment to be inoperable).
Calculate the radioactive release to the environment and resulting doses from the RER, containment spray syste=s, etc. over the 30-day period of operation.
State all of the assumptions that were used in your analysis and verify that they are conservative.
n Include the following parameters in your analysis:
U4.~972
\\
1-38
- Concentration (uc/ce) of iodine and noble gas a.
activity in the primary containment su=p water following a LOCA using the source ter=s specified in Regulatory Guide 1.7.
b.
Temperature curve vs. time for water being circulated thru pumps following a LOCA.
e.
Expected maximun leak rate (cc/hr) thru pumps seals, flanges, valves, etc.
d.
Partition factor for iodine.
Adsorption and filtration efficiencies of the filter e.
train used on the exhaust system for the engineered safety features area and whether t' ese systems meet the requirements of Regulatory Guide 1.52.
(2)
Provide an estimate of the total amount of leakage that could occur prior to isolation of failed equipment assumed (1) a pump seal failure and (2) a line severance several hours after the LOCA event.
E4-073
1-39 3.
Provide a table of data used for your loss of coolant accident analysis as illustrated in Table 15-2 of the
" Standard Format and Content of Safety Analysis Reports for Nuclear Power Plant," (Revision 1), October 1972.
(S A.'CY$
?
A.
.TIONAL TNFORMATION REQUIL ;ENTS
),gg
, THREE MILE ISLAND NUCLEAR STATION, UNIT 2 DOCKET NO.90-720 13-1/2.3.2 Provide information as to whether the wind speed values listed in Table 2 3 4 (pages 2.3-21 through 2.3-24 in the FSAR) are the upper, lower or mid-point values for the particular wind speed cate-gory represented.
13-2/P. 3. 2 Since Figures 2.3-14 through 2.3-21 in the FSAR may not be the same size in the reproduction as in the original drawing, the legend "Horiz. Scale 1" = 3000' " may not apply to these Figures in their present format.
Provide a horizontal scale for these Figures similar to the elevation scale.
13-3/2.3.2 Figure 2.3-22 lacks a distance scale.
In place of Figure 2.3-22, provide a larger scale topograph-ic map, showing site boundaries, plant structures and the location of the meteorological towers.
The map should include a distance scale and indi-cate the direction of true north.
13-4/2.3.3 Regulatory Guide 1.23 indicates that the lower temperature and wind sensors on the meteorlogical tower should be located at 10 meters (30-feet) above ground while the onsite towers have the lowest temperature sensors at the 25-foot level.
Provide either a proposed plan to realign the instrument array configuration to conform to the recommendations of Regulatory Guide 1.23 or a justification for the present configuration which deviates from that recommended in this Regulatory Guide.
13 4/2.3.3 Provide either a proposed plan of control room monitoring of wind speed and direction and tem-perature difference between two levels on the tower as recommended in Regulatory Guide 1.23 or a justification for not having such meteor-ological parameter monitoring capability in the control room.
13-5/2.3.3 State accuracy of instruments in Table 2.3-3, 11-6/2.3.4 Justify using a wind speed of 1.0 mph for calms.
13-7/2.3.4 State which wind speed categories in Table 2.3 h include calms.
13-8/2.3.5 Provide estimates of annual averace X/Q values for sixteen radial sectors to a distance of fifty miles in Table form.
WC7[i
. 3_n1 11-9/3.3.1 Show vertical velocity profile.
13-10/3.3.2 State the maximum rate of pressure drop that can be withstood by the Category I structures.
13-11/11.6 Provide justification for not having environmental monitoring stations within the river valley north and south of the plant in view of tae fact that wind flow may be channelled alone the river valley's axis.
13-12/15.0 Provide a table listing the X/Q values used in the calculations in this chapter.
13-13/16.0 Provide a description of the proposed post-oper-ational meteorological monitoring program, includ-ing tower location, instrumentation, data disposi-tion, and provisions for control room monitoring of meteorological parameters.
64'076
0TTIONAL INFORMATION REQU
- ENTS 7_g,'
THREE MILE ISLAND NUCLEAR STATION, UNIT 2 DOCKET NO.60-120 13-14/2.4.1 Hydrologic Descriution Provide a topographic map of the island which shows plant facilities, including the levee, and indicates plant drainage.
13-15/2.h.2 Floods Provide the Agnes flood levels recorded in the site area, and compare same to your " design flood" and levee profiles.
13-16 /2. h. 3 Probable Maximum Flood (PMF) on Streams & Rivers Compare PMF water level estimates with Agnes watar levels.
Explain the last sentence of Section 2.h.3.A.
' '-1 ' /2. h. o Channel Diversions Discuss the potential for seismic or flood diversion of the river at the site.
13-18/2.4.10 Floodine Protection Reauirements Describe all flood protection provisions.
Provide riprap gradation curves, and discuss its durability and maintenance.
Will the access bridge withstand a PMF?
13-19/2.4.11 Low Water Considerations Provide an instantaneous low flow frecuency curve, and an analysis which indicates the lowest flow and water level at which shutdown of both units can be maintained.
Provide your estimate of th? levas*
flow and level you believe possible.
Vinimum f'.ows of record in Sections 2.4.11 and 2.4.11.3 conflict.
What provisions are provided to prevent trash from blocking safetv-related pumps at all postulated river conditions?
(?4. v y/,
o 6
ENOLOSURE 2 ADDI'I'IONAL INFORMATION REQUIREMENTS THREE MILE ISLAND NUCLEAR STATION, UNIT 2 FINAL SAFETY ANALYSIS REPORT b4~C78
2 --I REQUEST FOR ADDITIONAL INFORuATION THREE MILE ISLA'ID NUCLEAR STATION, U'iIT 2 DOCKET No. 50-320 9-1/
For the evaluation of containment response following a design basis 6.2.1.3 LOCA, it is not apparent that the additional energy release from steam generators during post-reflood period has been included in the containment pressure response analyses.
We will need the results of the contain=ent pressure transient analyses for a spectrum of cold leg breaks.
Includa the effect of post blowdown energy sources, such as core stored energy and decay heat, primary system =etal stored energy, and steam generator stored energy.
The analyses should be extended through the initial blowdown, reflood, and post-reflood phases of the postulated accidents.
9-4/
For the break producing the highest containment pressure, provide 6.2.1.3 an energy distribution initially, at the end of blevdown, at the end of reflood period, at the time of containment peak pressure and post-reflood of the following energy inventories:
a.
the energy inventories in each component of the reactor coolant system, the steam generators, reactor coolant, steam generator fluid associated metals, and refueling water storage taak.
b.
the amount of energy from the sourcea including decay heat, zirconica water reaction and feedwater.
c.
the amou t of energy absorbed by each heat sink in the cor.tain=ent, containment atmosphere water, contain=ent atmosphere air, con-tainment structures and =iscellaneous heat sinks.
d.
the amount of energy removed by containment spray and residual heat exchangers.
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2-2 9-3/
The following questions relate to the conservativeness of the 6.2.1.3 assu=ptions used for determining the mass and energy releases for containment analysis.
Infor=ation should be provided assuming full ECCS operation and for mini =um ECCS operation.
Blowdown Provide an analysis of the = ass and energy for a spectrum of a.
hot leg and pump suction cold leg break sizes using methods and assunptions that are conservative for containment analysis.
For each break location, the following infor=ation should be provided:
1.
= ass release rate as a function of time; and 2.
energy release rate as a function of ti=e.
b.
Describe or reference the transition boiling correlation used to predict heat transfer from the core during blowdown.
Define the criteria that were used to establish the conservatism of this correlation.
c.
Provide the average core te=perature initially and at the end of blowdown.
d.
Describe how the pri=ary system volu=e which is used in calculating the initial liquid mass contained in the primary system is determined.
Provide the temperatura assu=ed in the ca'culation of the pri=ary system volu=e, and the assumed pressurizer water level.
Discuss the conservatism of these values from the standpoint of contain=ent analysis.
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2-3
- e.
For the =ost severe hot and cold leg breaks, provide the values of the heat trarsfer coefficients used in the stea generators.
Discuss how these values are conservative for contain=ent analysis.
f.
Provide analyses shewing the effect of node spacing and nucleate boiling heat transler coefficients on pri=ary =etal heat flow.
g.
Describe the =ethods used to calculate the initial core stored energy used in the containment pressure calculations.
Provide values of the initial and decay power level, gap conductivity, burnup, densification and fuel conductivity.
Discuss how these values are conservative for contain=ent analysis.
Reflood a.
Describe in more detail the model used to predict the = ass and energy release to the containment during the reflood period.
Discuss the conservatis: in the =odel with respect to
-,-14 4 :ing the energy release to the contain=ent.
Include the assu=ptiens
=ade regarding all energy sources, the flew resistance in the broken and intact loop, and the specific volume used in each flew ele =ent.
b.
Discuss the assumptions =ade in the reflooding calculation re-g arding steam condensation by the emergency injection water.
If condensation is assu=ed, provide justification based on applicable experf= ental data; 1.e.,
data corresponding to the conditions in the primary system.
Provide the results of a sensitivity analysis 64 0E1
2-4 showing the = ass and energy released and the effect on containment pressure if:
(1) no condensation and partial ECCS operation is assu=ed, (2) no condensation and full ECCS operation is assumed, (3) condensation and partial ECCS operation is assumed, and (4) condensation and full ECCS operation is assu=ed.
Discuss the assu=ptions =ade regarding separation of entrained c.
liquid leaving the core during the reflooding period.
We believe a conservative approach would be to assume no liquid separation so that all liquid leaving the core would enter the steam generator and be available for heat.
d.
Discuss the assu=ptions =ade in calculating the carryout fraction from the core (ratio of core axit flow to core inlet flow) during reflood.
These assu=ptions should be justified by comparison with the results of the FLECHT experiments for average core conditions during reflood.
We believe a carryout rate fraction of approximately 0.8 would occur until the 10-foot level in the core is covered.
Assuming mnrimum ECCS, provide the mass and energy release rates e.
as a function of time.
f.
Assuming minimum ECCS, provide the mass and energy release rates as a function of time.
Post Reflood In the post-reflood phase of the LOCA for cold leg breaks, when the a.
64.'~Cb2
2-5 core has been recovered with water, a two-phase mixture of steam and water will ba generated.
Provide an analysis showing height that the two-phase mixture will rise above the core.
If any water is calculated to enter the steam generator,' provide the = ass and energy release rates to the containment as a function of time.
b.
Describe in detail the analytical models used to calculate the additional =Isa and energy release to the containment during the post-reflood period after the core has been recovered with water.
c.
Provide a list of parameters used to determine the mass and energy release during the post-reflood period.
This list should include the void fraction and static liquid head in the downcomer, core, upper plenum and steam generator.
For the hot legs, steam generators and cold legs, provide the steam flow rate and pressure drop.
d.
Provide or reference all heat transfer and fluid flow correlations used in the analysis.
9-4/ state the minimum containment backpressure that has been ased in the 6.2 analysis of the emergency core cooling systenc.
Justify this value to be conservatively low by describing the conservatism used in the assumptions of operating containment conditions, modeling of the heat sinks, heat transfer coefficients to the heat sinks, heat sink surface area and any other parameter asst =ed in the analysis.
Provide the contain=ent pressure, te=perature and sump te=perature response for the most conservative assu=ptions.
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2-6 9-5/ Related to subco=part=ent pressure response analysis:
6.2.1.3 Describe the analytical model in det&II, assu=ptions and appropriate a.
bases, used in calculating the subco=partment pressure response.
b.
Describe the nodalization sensitivity studies performed to determine the mini =u= nu=ber of volume nodes required to concervatively predict the nav4m pressure for each subco=partment.
The nodali=ation sensitivity studies should include consideration of spacial pressure variation in the axial, radial, and circu=ferential directions.
Pro-vide sche =atic drawings of each subcompart=ent showing the general arrangement of the subco=part=ent structures, ce=ponents, piping, other =ajor obstructions, the nodalization, and the connecting flow paths between volu=e nodes.
Specify the nodal volumes, flow coefficients and flow areas usca to c.
calculate the flow between nodal volur.as.
This infor=ation should be of sufficient detail to allow confirmatory analyses to be perfoc=e.d.
d.
Clarify the =anner in which the flow coefficients were calculated.
Graphically show the pressure variation with ti=e for each sub-e.
compart=ent.
The maximum pressure nodal volume and other represent-ative nod' pressures should be included.
f.
Discuss the manner in which =ovable obstructions to vent flev (such as, insulation, ducting, plugs, and seals) were treated.
I.mclude analytical justification if credit is taken for the rc= oval of such ite=s to obtain vent area.
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2-7,
g.
Provide analyses of the pressure differentials for the pressuri:er enclosure and pipe annulas in the biological shield.
h.
Provide both the design pressure and the calculated transient pressure response for each of the subco=part=ents analyzed.
1.
Provide the = ass and energy blowdown rates in tabular for=, fro =
time :ero to abcut 1.5 seconds at approxi=ately 0.05 second inter-vals for the limiting case in each subco=part=ent.
9-6/ Information should be provided for a contain=ent pressure respense 6.2.1.3 analysis of stea: line and feedvater line break accidents to include the follewing:
a.
The = ass and energy release as a function of ti=e.
b.
The to.tal = ass and energy released into the contain=ent.
c.
A description of and justification of all the assu=ptiona used to predict the = ass and energy release.
d.
A detailed discussion of the analytical model used to predict blowdown flow rate and energy.
9-7/ Provide the =ethod and results of analysis of the jet forces used in 6.2.1.2 establishing the structural loads within the contain=ent.
9-S/ Provide the following information regarding the generation of hyd cgen 6. 2. 4. _3 following a design basis loss-of-coolant accident:
a.
Craphically show the Dategrated hydrogen production (f t ) within the contain=ent as a function of time for each of the potential sources of hydrogen; i.e., radiolytic deco = position of water, 5%
r,1rconiu=-water reaction, evolution of entrained hydrogen in Ed ~OS[5
2-8
, primary coolant and alu=inum corrosion, b.
Graphically shew the assu=ed corrosion rate of aluminum as a function of time.
A discussion of the magnitude of hydrogen off-gassing expected to c.
occur from zinein the paint primers and top coatings used in the containment and the analytical and experimental bases to support these expectations.
Include the source of hydrogen as applicable in the analysis requested above, d.
Provide a listing of the zine and aluminum components within the containment, as well as the mass and surface area of these co=ponents.
9-9/ A catalytic reco=biner is proposed for use in the hydrogen control 6.2.5 system.
The FSAR does not provide sufficient infor=ation related to the design, testing, and perfor=ance capability of the system.
Provide this infor=ation, together with a detailed discussion of the develop = ental and test program that was conducted to demonstrate the performance capability of the system.
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ENCLOSURE 3 ADDITIONAL INFORMATION REQUIREMENTS THREE MILE ISLAND NUCLEAR STATION, UNIT 2 FINAL SAFETY ANALYSIS REPORT
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Enclosurel 3-1 ADDITIONAL INFORMATION REQUIREMENTS THREE MILE ISLAhD liUCLEAR STATION, UiSIT 2 00CKET NO. 50-320 ENVIRONMENTAL REPORT A new, amended or supplemental Environmental Report for the Three tiile Island Nuclear Station, Unit 2, will be required to update the Environmental Report for Units 1 and 2, filed on December 10, 1971, as amended.
The updated Environmental Report should include data generated after that which was used in the initial Environmental Report, including but not limited to:
1.
Updated population data and projections 2.
The results of baseline and other environmental studies conducted during plant construction that were required by conditions in the FES.
3.
Plant design modifications and their related changes in effluent discharges with resultant impact on the environment.
4.
Revised need for power projection including discussion on the conservation of energy issue.
5.
Revised data on alternative power sources including the effects of the energy crisis on projected costs of alternative fuels.
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