ML19220B422
| ML19220B422 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 02/06/1976 |
| From: | Ross D Office of Nuclear Reactor Regulation |
| To: | Deyoung R Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 7904260012 | |
| Download: ML19220B422 (22) | |
Text
'
~
g FE5 u o kl6 f%
Docket Noc 50-320 j
- 7. C.3f.Youc $, Assistant Director for Light Water Re actors, '
^
THRE:: MILE ISLAND 2 SER INFUT Plant Name: Three Mile Island 2 Docket No.: 50-320 Milestone No.: 24-21 Licensing Stage: OL Responsible Branch & Project Manager: LWR 2-2; H. Silver Technical Review Branch Involved: Reactor Systeus Branch Description of Rev1w: SER Input Requested Completion Date: Nove=ber 19, 1975 Review Status: Co=plete The enclosed report contains the evaluation performed by the Reactor Systetas Branch on Three Mile Island 2.
This review included sections 1.5, 4.1, 4.4, 5.1, 5.2.2, 5.3, 5.5, 6.3.1, 6.3.2, 6.3.3, 6.3.4 and Chapter 15 frota the Standard Format Revision 1.
This report concludes the review of currently available information.
The results of the review by RSB of subsequent infctmation such as ECCS and Technical Specifications will be submitted as supplements ta this SER.
Similarly, the review of areas of the Three Mile Island 2 design which require further effort or documented conmitments on the part of the applicant will be reported as supplements. Thase areas are s e nrized as follows:
1.
Cot:mitment and schedule for confor=ance with requirements identified in NRC evaluation of RAW-10099. (ATWS) 2.
Demonstration that an oi.erator can reasonzbly be eroected to accomplishtthe evitchover of the ECCS from injection to re-circulation =edes following a LOCA.
3.
Redesign of cross-connect line between 3cv pressure injection trains to eliminate requirement for manual valve opening.
4.
Cot:nitment to conform to Regulatory Cuide 1.79 requirements for recirculation tests and core flooding tests or to provide acceptable alternative programs.
' W U.t 3 7904'260012
~
6 $75 F":
u
%.C.3eyevy
- We 5.
Resolution of various outstanding issues related to steam line breaks, feedvater line breaks and secondary systen desir.,
NIE
-3. F 'Ro Es f ".'1T 2Eg-M G
~~c s s y
-., Assistant Directer for Reactor Safety Division o' *~
S %:4tms ScEck
Enclosure:
SER input ce:
S. Encauer R. Heise:.an R. Boyd I. Knial H. Silver T. Novak W. Minners J. Watt DIS HIBUTION M File 50-310 SER. Reading RSB Reading
'p."R o s s l
e.
l
% /
TR:RS3 l
IR:RSB _
. JRlRSB DSS:Rs
- D TJat
._'4M1::n I'Jbvak
.._.[;
s 1l lIbE._. Al$ blI0-L\\ L1' - TIL_'lLS--
oeve
- Foran AEc.313 iRev. M)) AECM O240 N s. s. 2 3 v e *
- u t h ? **'*?**G
' F P tc Es t e74 928-8 se
THREE MII.E ISLAND 2 SAFETY EVALUATION REPORT 1.5 Rtcuire=ents for Further Technical Infor.ation The applicant has identified in Section 1.5 of the Three Mile Island 2 FSAR develop =ent progra=s applicable to the design.
These progra=s were initiated to establish the final design.
A "Once Through Stea= Generator Tast" was completed and reported in B&'J Topical Report 3Ati-10027. NRC approval was docu=ented in conjunction with a previcus 3&'J related Safety Evaluation.
Ther=al and Hydraulics programs were verified through a 1/6-scale
=odel flow test.
Test data analysis and docu=entation were sub=itted a: 3&'I Topical Report 3A'J-1003 7 (Rev. 2).
NRC approval was documented in conjunction with a previous 311 related Safety Evaluation.
4 P.
4 O' -.
.s)
I a
~
9 9 -
4.0 REACTCR 4.1 Su=marv Descrintion The.esign cf the Babcock and Wilcox pressurized water reactor for Three Mile Island 2 is similar to others fjavis-imsse 1 and Rancho Seco) recently approved for operation. The core consists of 177 fuel assemblies having 208 fuel rods each. The design heat output is 2772 MWt.
Full and part length control rods, dissolved boron, and burnable poison rod asse=blies are utill:ed for reactivity control.
4.4 Thermal and Hvdraulic Desien The Ther:al-Hydraulic design of the core for the Three Mile Island 2 plant was reviewed. The scope of the review included the design criteria and design basis as i=ple:ented in the final core des 1gn.
We have evaluated Three Mile Island 2 on the basis of a design power of 2772 MWt vith 13x15 fuel asse blies. As shown in Table 4.4-1, the ther=al and hydraulic design para eters for Three Mile Island 2 are identical to Rancho Seco and similar to Davis-3 esse 1.
The principal criterion for the ch artal-hydraulic design of a reactor is to prevent fuel rod da= age by providing adequate heat transfer to the various core heat generation patterns occurring during normal operations and anticipated transients. The applicant has denonstrated through the'use of the Westinghouse W-3 corrolation for departure from nucicate boiling heat flux ratio (DN3R) that a D:732 greater than 3 is maintained for steady-state design over power (112".)
~
and anticipated transient conditions.
a t) (*f[3
O 3
4 Table 4.4-1 Thermal-Hvdraulic Desi n Su==arv Ccecarisen of Three Mile Island 2. Davis-3 esse 1, and Rancho Seco Three Mile Davis Rancho Island 2 Besse Seco Design Core Heat Output, Wt 2772 2772 2772 Nominal Systes Pressure, psia 2200 2200 2200 0
Vessel Coolan: Inler Te=perature, F 557 355.4 557 7essel Coolan: Outlet Te:perature, F 607.7 608.6 607.7 Total Heat Transfer Surface Area in Core, f' 49734 49734 49734 Average Heat Flux, Stu/h-ft' 185,090 185,090 135,090 9
Maxi =u= Heat Flux, Stu/h-ft' 576,885 554,200 576,885 Average Ther:al Cutput, kW/ft 6.105
- 6. 105 6.105 Maximus Design Ther=al Output, kW/ft 19.03 13.28 19.03 o
Maxi =u= Cladding Surface Te:perature, F 654 654 654 Average Core Fuel Te:perature, F 1200 1200 1200 Maxi =um Fuel 7e:perature at Hot Spot, F
4170 4060 4170 6
Total Reactor Ccolant Flcw, 10 lb/h 137.8 131.32 137.3 Core Average Coolant Velocity, fps 16.52 15.74 16.52 DN3 Ratio at Design Overpcwer 1.39 1.41 1.39 DN3 Ratio at Design Power 1.75 1.79 1.75 F
/)=_
p*-
g v
9 9 The analyses indicate adequate argin over the 1.3 DN3R limit.
This permits the conservative assessment with 95". probability, at the 957.
confidence level, that the hot rod in the core will not experience a departure frc= nucleate boiling condition during nor:21 operation or transients that are anticipated to occur with =oderate frecuency.
The i= balance trip se points are established for the reactor pro-tection syste= functica to ensure that thermal design criterion on DNER and fuel ta=perature li:1:s vill not be exceeded. The i= balance limit envelope and the appropriate design conditiens will be provided in the Technical Specifica icns for Three Mile Island 2.
The analysis presented in the FSAR considered =axi=un design conditions and most prcbable condi:1cns. The for:er is the cost severe condition and was used to identify DN3R and fuel temperatures during nor=al and anticipated transients. The latter provides a convenient comparison for the expected conditions in the core.
It is expected that the limiting conditions for opera:1cn vill be identified by LCCA analysis to satisfy Apcendix K recuirenents.
The naximum design conditions of the FSAR could then be bounded by the Appendix K liciting conditions.
The Appendix K analysis and Technical Specifications will be reviewed by the staff when submitted. The results will be reported as supplements to this SER.
DuringtheDavis-3esserevie'bthestaffhadrequiredfurthercon-r r
-gr e o t) Gr
_. sideration of the effect of stuck open internal vent valves on the ther=al hydraulic design and core cooling characteristics.
~.4 has responded on a generic basis by submitting a report, "354 Operating Experience of Reactor Internal Vent Valves".
The staff has evaluated this report and concluded that a flow penalty due to internal vent valve leakage need not be applied. The applicant must, however, i=plement a program of inspection and tes of the valves at each re-fueling.
Single-1cep operatica while the reactor is critical is prohibited until the satisfactory completica of a single-loop test program.
This limita:1cn is to be noted in the Technical Specifications.
On the basis of cur review of the ther:al-hydraulic characteristics of Three Mile Island 2 and cc=parisons with Davis-Besse 1 and Rancho Seco, we conclude tha: the thermal hydraulic design of Three Mile Island 2 is acceptable.
ks (>m..(. s.
@ 5.0 RE. ACTOR CCCI A'.*T SYSTDi 5.1 Sn-,nry Descriotion The Three F.ile Island reactor coolant system consists of the reactor vessel, two vertical once-through steam generators, four shaft-sealed reactor coolant pumps, an electrica.'.ly heated pressurizer, and interconnecting piping.
In =ost important espects the system is similar to the Davis-3 esse and Rancho Seco systems. The stens generators on the Davis-Besse plant are physically located higher relative to the rest of the system than in this plant and Ranche Seco. We conclude that the overall design of Three Mlle Island 2 is acceptable.
5.2.2 overeressure Protection The pressure relief syste= prevents overpressurization of the reactor coolant pressure boundary under the most severe transients and limits the reactor presnura during nurnal operational transients. Ove rpressure protection for the reactor coolant pressure boundary is accomplished by utilizing the two safety valves.
These valves discharge to the pressurizer quench tank through a co==an header froc the pressurizer.
The reactor coolant systen (RCS) safety valves, in conjunction with the 22 steam generator safety valves, and the reactor protection system, will protect the RCS against overpressure in the even: of complete loss of heat sink.
The B&.? topical report on overpressure protection (3AU-10043) identifies the flow capacity safety margin as two for the pressurizer safety valves and 6% for the steam generator safety valves.
'F(1 -(12:Ol
While the staff's review of t.his report is'not comolete. the staff considers the report to have de=onstrated an adecuate relief capacity for the primary a id secondarv systens.
The peak RCS pressure follcwing the tors transien is limited to the ASME code allowable (110% of the design pressure) with no credit taken for operation of RCS relief valves, steam line relief valves, stes= du=p system, RCS pressurizer level control syste=, or pressuri:er spr y.
Four upsets were considered, control rod withdrawal, turbine trip, cc=plete loss of pcwer and loss of feedyater ficw. Conservative assu=ptiens were =ade for the analysis of each upset. All trip setpoints were assumed at =aximu colerances.
The control red withdrawal transient is more severe at zero or Icv pcwer producing a peak rimary pressure of about 25c5 psig which is 85 psi below the 110% li=1t.
Turbine trip from =axi=um overpower conditions produces the =os severe ec=bined pressure transient on both the s:ca: syste= and the reactor coolant syste=.
The stea syste pressure rising =c=entarily to within approxi=ately 15 psi of the 110% design pressure.
The staff concludes that the tsign of the pressure relief syste confor=s to the Co==ission's regulations and to applicable regulatory guides, staff technical positions and is acceptable.
_8 _
5.3 Thermal-Hydraulic Svstem Desizn The ther=al and hydraulic design bases are discussed in Section 4.4 5.5 Component and Subsystem Design 5.5.1 Reactor Coolant Pu=os and Motors The reactor coolant pe=p is designed to provide adequate core cooling flow and hence sufficient heat transfer to =aintain a DNSR of at least 1.3.
Sufficient pu=p rotational inertia is provided by the flywheel to provide continued flow following a loss of,pu=p power such that the reactor neutron power can be reduced before MiB limits are exceeded.
~
5.5.2 Stea= Generator The steam generator is a vertical, straight-tube-and-shell heat exchanger and produces superheated steam which is controlled to
=aintain a constant throttle pressure over the pcwer range.
The primary reactor coolant enters the steam generator hemispherical head and flows downward inside the tubes giving up heat to generate staas on the shell side secondary loop.
The tube and tube sheet boundary have the sa e design pressure and temperature as the rest of the pri=ary coolant boundary.
As the steam generators provide the heat sink for the primary reactor coolant system, the system is physically arranged to assure natural circulation for decay heat re= oval.
(
g ()
v% as
. O
_g_
5.5.7 Decav Heat Renoval Syste:
The decay heat removal syste: CpERS) is designed to remove decay heat and sensible heat frc: the RCS and core during the latter stages of cooldown. The system alsc provides auxiliary spray to the pressurizer for couplete depressurization, =aintains the reactor coolant te=perature during refueling, and provides the means for filling and draining the refueling cavity. In the event of a LCCA, the decay heat re= oval pumps are used for low pressure injection of borated water into the reactor vessel for emergency core cooling.
The decay heat renoval syste: is placed.into operation approximately six hours af ter initiation of plant shutdown when the temperature and pressure of the RCS are below 2S0 F and 260 psig, respectively.
Assu=ing that two pu=ps and coolers are in service, and that each cooler is supplied with component cooling wate at desi n flow and 3
temperature, the DERS is designed to reduce the RCS temperature to 140 F within 14 hours1.62037e-4 days <br />0.00389 hours <br />2.314815e-5 weeks <br />5.327e-6 months <br />.
If one of the two pumps er one the two coolers is not operable, safc cooldown of the plant is not co= pro =ised; however, the time required for cooldown is entended. The applicant has shown that, assu=ing only one train is available, the plant can be shut dcwn to below 212 F within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
The decay heat removal system for this plant was designed in accordance with Criterion 6 of the 1967 General Desian Criteria which' called for
"... reliable... decay heat renoval sys tens...." not specifically recuiring the application of single failura criteria.
The staff had expressed the concern that the single active failure of either of two isolation valves would prohibit. initiation of the RHR system during a normal shutdown. The applicant has provided
> r - f >, 3,3 iti
,q
assurance that there is an abundant inventory of water to continue cooling with the steam generators until any conceivable repair could be performed on the valves.
5.5.10 Pressurizer The pressurizer naintains the RCS pressure during steady-state operation and li=its pressure changes during transients.
It contains water volume, sized to provide the ability of the system to t
experience a reactor trip and not uncover the icw level sensors in the bottom head and to =aintain the pressure high enough so as not to activate the high pressure injection syste=; and a volu=e of steam, sized to provide the ability of the systen to experience a turbine trip and not cover the level sensor in the upper head.
Electric heater bundles, located in the lower section, and a water spray no::le in the upper section =aintain the stes: and water at the saturation te=perature which corresponds to the desired reactor coolant system pressure. During outsurges, as the RCS pressure decreases, so=e of the water flashes to stea: and the electric heaters restore the nor a1 operating pressure. During insurges, as RCS pressure increases, the water spray condenses steam to reduce
~
the pressure, to the nor:21 operating level. Two ASMI code safety valves are connected to the upper pressut
.er head to relieve systen overpressure. A pilot-operated relief valve is also provided to limit the lifting frequency of the code safety valves..
The safety and relief valves discharge to the pressurizer quench tank, located within containment.76-024
D 5.5.13 Saferv and Relief valves The pressuriser safety valves are bcllows sealed, balanced, spring-loaded safety valves which are provided wi:h a supple = ental back-pressure balancing piston for handling a bellows failure. The pressurizer relief valve is an electrically actuated, electrically centrolled, pilot operated, pressure leaded, relief valve.
The ec=bined capacity of the pressurizer safety valves is 690,000 lbs/hr at 2950 psig. The pressurizer relief valve capaci:y is 112,000 lb=/hr.
f.5.14 Internal vent valves The eight core support internal vent valves are located on a co==cn plane in the upper core support veld =ent above the outlet no::les.
These valves provide a direct flow path between the upper core region cnd the inlet annulus in the event of a loss-of-coolant accident from an inlet (cold leg) line break. This ficw path provides for pressure equalization by the venting of steam to the break and petsits the emerge.cy coolant water to reflood the core rapidly.
5.5.15 Loose Parts Monitoring System Vibration and loose parts in the reactor coolant systen are =cnitored with a co=prehensive syste= designed to analy:e vibration data and permit accurate accessment of the status of the system.
The vibraticn signature analysis of the system warns of changes in components.
This permits cohponent removal from service for
~
inspection and/or repair before breakdown or damage occurs.
The system also monitors vibration of the reactor core, r( k)-_ b1ay g)-
so
@ 6.3 Imerzenev Core Coolinz Svstem 6.3.1 Desi n Bases The applicant states that the principal design basis for providing protection over the entire spectrum of break sizes for the ECCS is General Design Criterion 35.
Very s=all breaks that do not actuate the engineered safety features code of operation are acco==odated by the nor al =akeup cystes in confor:ance with General Desi;n Criterion 33.
Separate and independent flow paths are provided in the ECCS.
Redundancy in active co=ponents ensures that the recuired functions will be perfor ed if a single failure occurs. Separate energency pcwer sources are supplied to.he redundant active co=penents.
Separate instrument channels are provided to activate the syste=s.
The range of coolant system ruptures and leaks the ECCS is aesigned tc nitigate extends from leaks in excess of the capability of the reactor coolant makeup system up to cnd including a break ecuivalent in area to a double-ended rupture of the largest pipe in the reactor coolant syste=.
6.3.2 Systes Desi n The ECCS for Three Mile Island 2 consists of tae core flooding tanks (CFT), high pressure injection (RPI) system, and a low pressure injection (LPI) system. The HPI and LPI systers take water from a borated water supply tank (SWST) during the irjection phase fellowing a LCCA.
As the BWST supply nears depletion, the operator must, (on a low level alarn) realign the systen to a recirculation
- g ya r g.
$ )
Lo t; U
O O =cde taking water frem the reactor building su=p.
One open ites relative to the ECCS is proof by the applicant that an operator can reasonably be expected to =ake the transition from the injection to recirculation =cdes following a LOCA. As failure to acco plish this changeover terminates all core cooling, the staff has requested the applicant to provide a detailed sequential description of this operatica.
A sacend open ite: involves the capability of the ECCS to accceplish its fut...:1cn folicwing a break in a low pressure injection line concurrent with the single failure of the unaf fected injection train diesel. The applican~ is required to revise the " cross-connect" between the Vo low pressure injection trains into a configuration consistent with plants acv being licensed. Manual opening of valhes in the event of a need for train interconnection is not acceptable.
Technical Specifications are required relative to settings, li=its and surveillance of ECCS ele:ents including BWST levels and alar =s.
To minimize the potential for water ha==er occurring due to ECCS inject _on into dry lines, the applicant has stated that during nor=al operation the ECCS lines vill be maintained full. The staff requires that the capability to =aintain filled ECCS piping be observable prior to startup and that the venting provision constitute a periodic surveillance requiremer.t. Both require ents
=ust be identified in the Three Mile Island 2 Technical Specifications.
,. c.- gr o vt
( t) a4/
@ 6.3.3 Perfor:ance Evaluation The applicant has co=mitted to provide an analysis of ECCS perfor=ance for Three Mile Island 2 in accordance with 10 CFR Part 30 Para;raph 46 and Appendix K.
In addition to the revised LCCA analysis, the applicant is also ec=mitted to submit additional infor=ation in the specific areas of sinimus centainment pressure, single failure criterion, effects af boron precipitation on long-ter=
cooling capability, and sub=erged valves within containment. The adequacy of ECCS perfor=ance and the staf f's evaluation of the applicant's evaluation model will be reported in a supple ent to
~
this Safety Evaluation Report.
6.3.4 Tests and Insoections The applicant vill denenstrate the operability of the ECCS by subjecting all. systems and co ponents thereof to preoperational tests, periodic :esting, and in-cervice testing and inspect' ions.
The applicant has co==itted to the implementation of the bulk of the Regulatt ry Guides applicable to initial test progra=s.
In some cases, the applicant has provided accaptable alternates or justification for deviations. Relative to Regulatory Guide 1.79 the applicant is co=sitted to provide additional information relative to containment sump testing and core flooding flev rate to either justify alternate approcches or to document that the requirements will be met.
The staff vill review the Technical Specifications to determine the adequacy of periodic testing, in-service testing and inspections.
t ()--Nv'a v
o n r2 The results will be included in a supplenent to this SER.
Su==arv of ECCS Evaluation On the basis of cur evaluation we have concluded that for the ICCS design of Three Mile Island 2 to be acceptable the following is required:
1.
The applicant =ust demonstrate that adequate time is available following the icw level alars to reasonably expect the opera:or to respond and realign the ECCS fram the injection to the recirculation modes.
2.
The applicant is required to revise the cross-cennect be: ween the two low pressure injection ::ains so that an abundant supply of cooling water is assured following a LCCA in the injectica line and the single failure of the diesel powering the unaffected injection train.
3.
A satisfactory analysis of ECCS performance in accordance with 10 CFR Part 50 paragraph 46 and Appendix K is required.
4 Outstanding issues relativa to Regulatory Guide 1.79 relative to containment su=p testing and core flooding flow rate require resolution.
5.
Standard Technical Specifica:1cns will be applied to Three Mile Island 2.
When ready, the specification will be reviewed by the staff specifically for this plant.
Items currently identified include:
a.
venting and surveillance of ECCS lines; b.
borated water supply tank level limits for operatica, icw level alarm Icvel limits and surveillance procedures; and s ly b2b) c.
schedule for de=enstration of operability of local =anual handwheel backup for each ECCS valve.
e 4
e m
76 030
O.
15.0 ACCIDES*! ANALYS!S 15.1 General The safety analysis subcitted evaluates the ability of Three Mile Island 2 to operate without undue hazard to the health and safety of the public. The events pertinent to safe y as discussed herain are divided into twe groups; abnormal transients and postulated accidents.
15.2 Abnormal Transients The criterion, adopted to assure that the reactor coolant pressure boundary integrity is =aintained, is that the system pressure shall remain below the code pressure 11:1:s set forth in ASME Code Section III (110% of RCS design pressure). The criterion adopted to ensure that no fuel da= age has occurred is that the D:3R tus be greater than 1.30 throughout the transient.
The applicant has submitted analyses of abnor=al transients and has shown that the integrity of the RCS houndary has been maintained and that the mini =us DNER exceeded 1.30 for all analyzed transients.
The pressure transient which produced the highest reactor coolant system pressure was identified (3A'J-10043) as the control red withdrawal at low power conditions, resulting in a peak RCS pressure of about 2665 psig. The most severe secondary side pressure transient was the turbine trip from overpower conditions, resulting in a =aximum steam generator pressure of about 1140 psig.
'/{y (o\\ ') 4 c
O
@ The applicant has referenced 3A'a'-10099 (Reference 3) as their position regarding design features to =ake tolerable the con-sequences of failure to scran during anticipated transients.
The fiRC staff has completed its evaluation of this repcrt and identified the changes required. The applicant is required to provide a co==itzent and schedule for conformance.
The co=puter code "FCiER M I:i" used for several abnor al transients in the FSAR, is currently under review by the staff. Should modifications to the code be required, the effect of these changes on the Three Mile Island 2 analyses mus: be considered.
The evaluation of abnormal transients indicated that the transients presented do not lead to unacceptable consequences and are acceptable for issuance of the operating license.
15.3 Accidents The applicant has evaluated a broad spectrum of accidents that night result from postulated failures of equipment, or other inproper operation. These highly unlikely accidents have been analyced in detail by the applicant and are representative of the spectr im of types and physical locations of postulated events involving the various engineered safety feature syste=s.
As neced in Section 6.3, the Loss-of-Coolant Accident analysis for Three Mile Island 2 will be reviewed separately when received from the applicant.
The locked rotor accident was analyced by postulating an instantaneous seizure of one. reactor coolant pump.
Reactor coolant flow decreased n
s
@ rapidly and reactor' trip occurred as a result of high power-to-flow signal. The sinimum DNBR (1.3) was exceeded for less than two seconds. Cladding surface te=perature was calculated to reach 1800 F at the het spots on.57* of the rods. No fuel cladding failure or significant :irconium-water reaction would be expected during such a short period.
If the conservative assu=ption that the limited percentage of fuel would fail when the DN3R 11=1 is exceeded, the radiological consequences of this event would be bounded by other events considered.
Various issues have arisen relative to loss-of-secondary-coolant accidents (steam line breaks and feedwater line breaks). The applicant has sgr ed to analyze a spectrum of steas line breaks inside of containment to identify worst sir.gle active failures affecting core behavior and contain=ent integrity and to propose systas or procedural changes if required. Additional analpsis of a feedwater line break presenteu in resp nst to question S-3-21.49 is required to clarify and resolve the design and function of the
=ain stea=. solation valves.
The adequacy of the analysis and possible changes in the secondary systems will be reported in c supplement to this Safety Evaluation Report.
References 1.
3A'i-lC099, "35W Anticipated Transients Without Scram Analysis" Dece=ber, 1974.
2.
" Status Report on Anticipated Transiente tiithout Scras for 3abcock
& Wilcox Reactor" NRC December 9,1975.
3.
Letter from Karl Kniel to R. C. Arnold, November 21, 1975.
e ep P