ML19317E120
| ML19317E120 | |
| Person / Time | |
|---|---|
| Site: | Oconee  |
| Issue date: | 09/08/1966 |
| From: | Grimes B US ATOMIC ENERGY COMMISSION (AEC) |
| To: | Boyd R US ATOMIC ENERGY COMMISSION (AEC) |
| References | |
| NUDOCS 7912130831 | |
| Download: ML19317E120 (7) | |
Text
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L'NITED STATES GOVERN.\\!ENT Memorandum 10 THE FILES DATEL g{p g 1966
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THRU:
Roger S. 3oyd, Chief Research &, Power Reactor Safety Branch, DRL
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FROM
- 3. Crimes Research & Power Reactor Safety aranch, DRL
SUBJECT:
MEETING WITH DUKE PO'.iER COMPANY ON FORTHCOMING CONSTRUCTION PER' TIT APPLICATION 5 Q ".2G 9 On August 25, 1966, a meeting was held in the 3ethesda offices with representatives of Duke Power Co=oany, Babcock & Wilcox Company and Bechtel Corporation to discuss Duke's forthcoming application for a construction permit for the Keowee River site in North Carolina.
Duke will be its own architect-engineer but has hired Bechtel as a consultant on the prestressed, post-tensioned containment desi;:n and other areas.
Sabcock and Wilcox will provide the nuclear steam supply system.
Attendance at the meeting included the following:
M.
Mann AEC-REG E. G. Case AEC-DRL R. S. Boyd AEC-DP1 D. R. Muller AEC-DRL S. Grices AEC-DRL C. Long AEC-DRL l
R. L. Uaterfield AEC-DRL tl. C. Seidle AEC-CO Gene Watkins Duke R. L. Dick Duke-Constr.
l L. C. Dail Duke-Engr.
E. C. Fiss Duke-Engr.
Roy 3. Snapp Duke ('. lash.)
D. S. Robbins Duke Engr.
T. F. 'Jyke Duke Engr.
W. H. Owen Duke Engr.
P. H. Barton Duke-Steam Production S. E. Nabow Duke-Steam Production A. C. Thies Duke-Steam Production D. W. Montgomery B&W - Project Manager R. E. '.la s ch e r B&W - Engr.
'i. S. Lee Duke VP for EUGR C. D. Stratton Bechtel Corporation N. F. Rau 3echtel Corporation i
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Bay U.S. Savings Bonds Regularly on the Payroll Savings Plan 7 912130 D/
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THE FILES Date:
r Duke ?over Ccmpany representatives stated that the present schedule called for an applicatien for construction permit to be submitted about Decer.ber
- 1. 1966.
Future schedule dates tre 2s folicus: break ground ?! arch,1"67, i
pour concrete, September,1957; criticality, December 1979: en line, "ay, l
1971.
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i Tentative plans call for the second unit to follow apnrontrately one year later and, if convenient, Uni: 1 fuci vill be used in the Unit 2 startun to try to achieve an equilibriun core socner than otheriise possible.
The staff requested the applicant to lock at the safety aspects of this cro-i posal. Two safety aspects are presently suggested to my r.ind:
(1) the startup of an untried reactor syste= vith fuel which cenrains fission pro-l ducts and (2) the question of whether any unique core ohysica situations vould be involvad.
The applicant stated taa: the cen:2inrent dast;n iculd he based en the
" stretch' capacity of 2568 3Gt but : Sat therral analysis would Se based on the initial operating value of 2'02
'f":.
Reactivity transients vill he studied using the icver value.
1 The folleeing coints vere brought cut ui:h respect to the sita.
I (1)
The pla.c *:.11 he built 10 feet beleu the level of a lake to l
he backed un behird the planned Zacvec dan.
Ficoding (in case of break-i age of the earth dam) vill have to be censidered.
Cooling vater vill be aken fren the lake above a second dan and discharged to the lower lake.
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l (2)
The Isovee dam is subject to aporoval by the Federal Power i
Commission and if cereission is denied, the reacter would Save to be i
built at a di?ferant site.
(3)
Suke will have a progran designed to obtain ground level retaorolo:7 at the site.
(It apoears that this data might not be representative until the laka is es:ablished).
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(4)
Eter:ency station power wculd be obtained through an under-
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- round line frca a hydro plant to be built with
- Se dam. The hydro plant containing rio turbine-generators would be used for peaking loads and when not en line could be started in about one ninute.
There would I
be occasions when this pcver would not be availabic--every few years the penstock must be "devatered" and inspected.
(This could perhaps he sc1ved by having a s.agle diesel on site and running it centinually during the cutage).
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i THE FILES Date:
l 3abcock and '.;ilcox made a presentation which cutlined the differences between the Duke plants and the reactor described in the submittal (OMi-293 and supplement) on which a preliminary reviev was performed by the regulatory staff. The several dif ferences are listed in the table below.
l Duke B&L' c relin.
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(1) Higher power level 822 :Ne 750 F.le (2) Operating closar to thermal i
limits: ocerating pcwer 7% increase i
overpever 4% increase operating power D ;3R (3A'i-168) 1.55 1.60 6
6 (3)
Eigher coolant flow 131.3x10 1b/hr 118.3::19 1b/hr (4)
Increase in nunber of pins per assembly 209 200 L
(5)
Increase in length of active fuel 144 in.
100 in.
(6)
Increase in te=peratures :
0 inlet 550cF 540 F I
outlet 603 F 591 F 0
C (7)
Steam 2cnerator (still 350F superheat), steam pressure 925 psia 325 psia C
steam temperature 570 F 553 F j
Other differences from the original 3&W proposal include:
(1) The plants will share part of the en2ineered safeguards.
3edun-dancy in each set of safeguards would be provided by a single " swing' unit which could be valved to either plant.
i (2)
Each containment vill be prestressed, post-tensioned and have an incarnal
. lune of about 2 x 106 cubic feet. A non-insulated liner is planned.
The containment will rest on a slab base with no rock anchors.
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-(3)
A corr.on spent fuel pit will be provided.
(4)
The units will have a 25% bypass capability but will have tha capability to dung a 100% load by venting secondatf steam to the atmosphere.
(5)
About 90% of the safeguards will be outside the containment.
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I THE FIL25 Date:
i The staff listed a nunber of itets which should be more fully treated (than in the OU.! pralininary sub=ittal) in the hazards analysis to be prepared by Duke.
These iters had been previously discussed with OEl.
i A. Containnent (1)
Sufficient information should be included in the apolication l'
to demonstrate how the design criteria vill be met.
For exampic, containnent isolatien criteria should be defined and v2lve arran;erents for each type of penetration should be supplied j
to shes how the criteria are cet.
(2)
Vessel classificaticns should be supplied as well as missila protection criteria.
i (3)
It was stated by the staf f that the design basis accident for the containnent should assume the fission product release and the metalsvatar reaction associated vith a core neltdown.
3.
Reactor Svsters (1)
Complete process and instruecntation diatrans should Se supnlied for all sys tems associated with the reactor, includine connenent cooling and service water systems.
i (2)
The design bases for the pri=ary system safety valves sSculd be provided in terms of load rejection capability and also analyses j
to show that these are adequate to protect the plant.
i (3)
A full description of the steam generators vill be required, including design bases and experinental data which would sup-port the once-through d0 sign.
j (4)
Information on the rod drives desired by the staf f includes i
(a) hcw the disenga;ecent of the nutator gears will be in-sured when a scram is required, (b) hev the de-energi:ation j
of the magnet cotts vill be insured and (c) what the maxin'in i
capabilitias of the drive are to acvc against physical rastrie-tions in the channels.
(5)
The methods of deternining (censuring) horon worth and rod worth during operation should be elaborated.
C.
Thermal Desinn A sensitivity analysis should be provided for the thermal design, shcuin; the effect of variation of important parameters on the probability of fuel burnout. A conparison of the 3AW-163 correlation with other current i
correlations shculd also be orovided. A study such as was presented in l
the preliminary analysis is satis factory.
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l THE FILES Date:
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i 3.
~.nnineered Safequards i
(1)
Separate process and instrunentation diagrars should be supplied for each engineered safeguard systen including all systens required for effective operation.
The systees should be detailed to the extent of a preliminary design.
The injection systens provided should be designed to prevent clad celting or allow only a small amount of melting af ter sny coolant loss accident with only encr-eency power available.
The break sines considered should rance i
f rom small leaks to the design basis break for the containment.
(2) All en2ineered safeguards systees which would transport radio-active gases or liquid outside the containment during an acci-dent should be evaluated for potential leakage oaths to the 4
environment.
(3)
Energency power ecpabilities should be defined including the minimum emergency power available durine operation of one or both niants.
(4)
!!inimum encineered safeguards available during operation of one or both plants should be specified, including the decree of l
redundancy to be availabic while narts of tbc syste= are under-j going maintenance.
E.
Accident Analysis J
(1)
The rod ejection accident should be analy:cd to shew that no prieary system damage would result.
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nodel used in reacti-vity accidents snould be analyzed to %,r=ine its sensitivity to variation of all important parameLetc. including Dopnier and coderator coefficients.
(2)
Accidental dilution of the primary systen from all sources of unberatad water should be analyzed.
(3)
Misa!1e danage from potential turbine f ailure cdes should be evaluated to shew that containnent, engineered safecuards, and systers required for a saf S shutdown vould be af fected.
(1). The need for isolation valves in the secondary system to cone vith postulated steam line brcsh accidents should be evaluated.
(3) An analysis of staan generator :ube n2ptures shoula be provided including methods for control, iteps in an orderly shutdown need j
for' isolation valves, 2nd resulting envirenrent hazards.
F.
Dual Reactor Installation Systems which are to be ccanen to both reactors should be clear 1v defined and the safety-related incaractions of each sys:2 avaluated.
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T:is FIL2S Date:
A discussion of the reliance en cross-connections to provide redun-dancy in the engine 2 red safeguards should be presented.
C.
Instrurentation and Control (1)
Test data relating to the in-core instrumentation research and development prograc should be provided.
Discussion accomeanyin' the data should encompass the ite=s which 35*1 has indicated as included in its research and develeprent program.
(2)
Psrticular attention should be ;iven to the safety systen desien to ensure that no single ciectrical, rechanical or hydraulic failure can prevent automatic safety action, initiation of en-gineered saferuards or containrent isolation.
Live shorte at the pcVer buses should be censidered.
(3)
Sufficient instrueentation sche: aries should be included to allow a determination that the pecposed systems conform to the Cen-mission's proposed criteria dated November 22, 1965, numbers 13 and 16.
(4)
Consideration should be given to separatinc the nuclear flux servo and the (level) safety systens.
(5)
Redundancy should extend to all aspects of the power /flou safety channels including devices which reasure and conpare power and flov signals.
(6)
In order to evaluate the " ultimate capability" of this plant to withstand serious accidents provide the folleving analyses:
(a) Assume that all rods are simultaneously vithdrawn under cold, clean conditions. The nodel may assure that the nucicar flun level or the high pressure trips raaet as des 12ned.
(b)
Discuss the adequacy of one boiler to remove heat fol-lowing a total (external and internal) s.c. blackout.
(c) Discuss the nanual accessibility of essential breakers under the assumption that, coincident with the desien basis accident, the d.c. voluage supply is lost.
(7)
Consideration should be given to the use of stored-energy cen-tainment isalation ralves actuated by f ail uafe (in the event of voltage loss) instrumentation.
T115 FILES Date:
(B)
Discuss the design features which assure that the control rods cannot be drawn beyond a safe maxi =um speed.
For exanole, if electronic oscillators are being used, what ensures that a fault in the oscillator would not allow its output frequency (and, therefore, the rod speed) to exceed safe limits?
cc.
E. G. Case C. G. Long D. F. Sullivan DP1 Reading RSPRS3 Reading
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