ML19261D721
| ML19261D721 | |
| Person / Time | |
|---|---|
| Site: | Crane, Davis Besse  |
| Issue date: | 05/09/1979 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| References | |
| NUDOCS 7906260183 | |
| Download: ML19261D721 (73) | |
Text
.__ _
5 50-320 Miscellaneous documents from NRR files dealing with Davis-Bessie as it relates to TMI2/14/78 - 5/9/79 2310 202
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DDonoghue,ADM DRR:Subj TRehm,ED0 EShomaker El JFelton, ADM DGrimsley, ADM The Honorable Morris K. Udall, Chairman CA, CKammerer Subcomittee on Energy and the Environment JMaynard,~ ELD Comittee on Interior and Insular Affairs JFouchard, PA U. S. House of Representatives BShiel ds, 0GC Washington, D. C.
2D515 s HDenton'; fiRR JDavis IE
Dear Mr. Chaiman:
Er'fr,EDOM.I-This is in reference to a telephone call to Ke Corne I M Dr. Henry Myers on April 18, 1979, NRC records regarding the Davis Besse Nuclear Reactor.in which he requeste Dr.
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indicated a special interest in any memoranda written by Re III's inspectors, and reports on the reactor.
I am pleased to provide you the enclosed records which relate to the September 24, 1977 transient at the Davis Besse Nuclear Reactor.
Attached to this letter is a list of these records.
Additional material is presently being compiled and will be provided to you in a few days.
Please call us if we can be of further assistance in this matter.
Sincerely,
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Lee v. G
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Execu::..
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Enclosures:
As listed in attachment 2310 203 CA 4/ /79 4/27/79 (Retvoed in EDO office)
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i ATTACHMENT 1.
Memo dated January 19, 1979, Keppler to Moseley and Thornburg (with three enclosures).
Recommendations For Notification of Licensing Boards and Request For Technical Assistance.
2.
Memo dated Febuary 28, 1979, Moseley to Thompson (with enclosure)
Notification of Licensing Boards (AITS F30468H2).
3.
Memo dated March 1,1979, Thompson to Vassallo (with enclosure - 2)
Information For Board Notification - Davis-Besse Units 2 & 3 and Midland Units 1 & 2.
4.
Memo dated March 6,1979, Vassallo to Christenbury, oroviding Item 3 above to Hearing Division Director and Chief Counsel, OELD.
5.
Memo dated March 7,1979, Moseley to Thompson, Notification of Licensing Boards (Discussion and Evaluation will be sent later).
6.
Memo dated March 12, 1979, Thompson to Vassallo transmitting Item 4 above.
7.
Memo dated March 28, 1979, Moseley to Thompson providing Evaluation of Concerns.
S.
Memo dated March 29, 1979, Moseley to Thompson acvising that Evaluation of Concern previously provided may change due to TMI-2 incident.
9.
Memo dated March 29, 1979, Thompson to Vassallo forwarding Evaluation of Concerns for transmittal to Boards.
- 10. Letter dated March 29, 1979, Scinto to Service List, re:
Davis-Besse, Erie, Greene County, Midland 1 & 2, Pebble Springs, and Three Mile Island 2.
- 11. Chronology of Turbine and Feedwater Pump Trips From Startup Test Report, dated February 8,1979.
- 12. Event Chronology for Rancho Seco, Undated
- 13. Chronology of Turbine and Feedwater Pump Trips From Startup Test Report, dated February 8,1979.
- 14. Turbine Trips & Feedwater System Problems Crystal River Unit No. 3
- 15. Events at B&W Facilities Similar in Nature to TMI-2 Event Which Deviated from Expected Teransient.
- 16. Additional TMI-2 Chronology from Monthly Operating Reports.
- 17. TMI-2 Chronology of Personnel Related Events.
2310 204 N2s O
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O HE40R.8J.*DUM POR: Leonard Bickvit Ceneral Counsel FROM:
Harold R. Denton, Director Of fice of Nuclear Reactor Regulation
SUBJECT:
PROPOSED OF.DERS Enclosed are draf t copies of two proposed orders which will confirm the co ittents of the Toledo Edison Company p.nd the Florids Power Corporation to shut down and nodify their facilities and operating procedures. Also enclosed are copics of related correspondence f rom the licensees.
Odzir.3t syned EI' E. G. Crt g
. Harold R. Denton, Director
" Office of Nuclear Reactor Regulation
Enclosures:
/a stated DISTRIBUTION Central Files HRDenton ECCase RReid RIngram TEngelhardt GCunn ingl.ac Dross 6
Contact:
Edson G. Case NRR s
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_ODI 318 (9-76) NRCM 0:40 U us o. oovsan =amt pathrine or rec a, isre - ese-eaa K
7590-01 UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION
)
in the Matter of
)
THE TOLEDO EDISON COMPANY AND
)
THE CLEVELAND ELECTRIC ILLUMINATING
)
Docket No. 50-346
)
COMPANY
)
)
Davis-Besse Nuclear Power Station,
)
Unit No. I ORDER I.
Tne T leco Edison Company (TECO) and The Cleveland Electric Illuminating Company (the licensees), are holders of Facility Operating License No.
NPF-3 which authorizes the operation of the nuclear power reactor known as Davis-Besse Nuclear Power Station, Unit No.1 (the facility or Davis-Besse 1), at steady state power levels not in excess of 2772 megawatts thermal (rated power). The facility is a Babcock & Wilcox (B&W) designed pressurized water reactor (PWR) located at the licensees' site in Ottawa, County, Ohio.
II.
In the course of its evaluation to date of the accident at the Three Mile Island Unit No. 2 facility, which utilizes a B&W designed PWR, the Nuclear Regulatory Commission staff has ascertained that B&W designed
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2310 206 e
7590-01 2-reactors appear to be unusually sensitive to certain off-normal transient conditions originating in the secondary system.
The features of the B&W design that contribute to this sensitivity are:
(1) design of the steam generators to operate with relatively small licuid volumes in the second-ary side; (2) the lack of direct initiation of reactor trip upon the cccur-rence of off-normal conditions in the feedwater system; (3) reliance on an integrated control system (ICS) to autonatically regulate feecwater flow; (4) actuation before reactor trip of a pilot-operated relief valve on the primary system pressurizer (which, if the valve sticks open, can aggravate the event); and (5) a low steam generator elevation (relative to the reactor vessel) which provides a smaller driving head for natural circulation.*
r Because of these features, B&W designed reactors place acre reliance on the reliability and performance characteristics o# the auxiliary feed-water system, the ICS, and the emergency core cooling system (ECCS) per-formance to recover from frequent anticipated transients, such as loss of offsite power and loss of normal feedwater, than do other PWR designs.
This, in turn, places a large burden on the plant operators in the event of off-normal system behavior during such anticipated transients.
"It is noted that although features nunbers 3 and 5 do not apply to Davis-Besse I to the same extent as they apply to other currently licensed 33W designed reactors, the other features are fully appli-cable.
2310 20L7 9
7590-01 1 As a result of a preliminary review of the Three Mile Island Unit No.
2 accident chronology, the NRC staff initially identified several human errors that occurred during the accident and contributed significantly to its severity. All holders of operating licenses were subsequently instructed to take a number of immediate actions to avoid repetition of these errors, in accordance with bulletins issued by the Commission's Office of Inspection and Enforcement (IE).
In addition, the NRC staff began an immediate reevaluation of the design features of B&W reactors to determine whether additional safety corrections or improvements were necessary with respect to these reactors. This evaluation involved numerous meetings with B&W and certain of the affected licensees.
The evaluation identified design features as discussed above which indicated that B&W designed reactors are unusually sensitive to certain off-normal transient conditions originating in the secondary system.
As a result, an additional bulletin was issued by IE which instructed holders of operating licenses for B&W reactors to take further actions, including immediate changes to decrease the reactor hich pressure trip point and increase the pressurizer pilot-operated relief valve setting.
Also, as a result of this evaluation, the NRC staff identified certain other safety concerns that warranted additional short-term design and procedural changes at operating facilities having B&W designed reactors.
2310 208 ew
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7590-01 These were identified as items (a) through (e) on page 1-7 of the Office of Nuclear Reactor Regulation Status Report to the Commission of April 25,1979.
Af ter a series of discussions between the NRC staff and the licensees concerning possible design modifications and changes in operating pro-cedures, the licensees agreed in letters cated April 27 and itay 4,1979, tc implement pecnptly the following actions:
(a) Review all aspects of the safety grade auxiliary feed ater system to further upgrade components for added reliability and performance. Present modifications will include the addition of dynamic braking on the auxiliary feedpump turbine speed changer and provision of means for control room verification of the auxiliary feedwater flow to the steam generators. This means of verification will be provided for one steam generator prior to startup from the present ma.intenance outage and for the other steam generator as soon as vendor-supplied equipment is available (estimated date is June 1, 1979).
In addition, the licensees will review and verify the adequacy of the auxiliary feedwater system capacity.
(b) Revise operating procedures as necessary to eliminate the option of using the Integratea Control System as a backup means for controlling auxiliary feedwater flow.
2310 209
- ~
7590-01 5-(c)
Implement a hard-wired control-grade reactor trip that would be actuated on loss of main feedwater and/or turbine trip.
(d) Complete analyses for potential small breaks and develop and implement operating instructions to define operator action.
(e) All licensed reactor operators and senior reactor operators will have completed the Three Mile Island Unit No. 2 simulator training at B&W.
(f) Submit a reevaluation of the TECO analysis of the need for automatic or administrative control of steam generator level setpoints during auxiliary feedwater system operation, previously submitted by TECO letter of December 22, 1978, in light of the Three Mile Island Unit No. 2 incident.
(g) Submit a review of the previous TECO evaluation of the September 24, 1977 event involving equipment problems and depress-urization of the primary system at Davis-Besse 1 in light of the Three Mile Island Unit No. 2 incident.
In its letters the licensees also stated that the actions listed in (a) through (g) above would, except as noted in item (a), be completed prior to startup from the current maintenance outage.
2310 210 e
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6-In addition to these modifications to be implemented promptly, the licensees have also proposed to carry out certain additional long-term modifications to further enhance the capability and reliability of the reactor to re-spond to various transient events. These are:
- The licensees will continue to review performance of tne auxiliary feec-water system for assurance of reliability and performance.
- Tne licensees will submit a f ailure mode and ef fects analysis of the ICS to the NRC staff as soon as practicable. The licensees stated that this analysis is now underway with high priority by B&W.
- The reactor trip following loss of main feedwater and/or trip of the turDine to be installed promptly pursuant to this Order will thereaf ter be upgraded so that the components are safety The licensees will submit this design to the NRC staff grade.
for review.
- Continued attention will be given to transient analysis and procedures for management of small breaks.
- The licensees will continue reactor operator. training and drilling of response procedures to assure a high state of preparedness.
2310 211 O
e e
7590-01.
The Commission has concluded that the prompt actions set forth as (a) through (g) above are necessary to provide added reliability to the reactor system to respond safely to feedwater transients and should be confirmed by a Commission order.
The Ccomission finds that operation of Davis-Besse 1 should not be re-suced until the actions described in paragraphs (a) through (g) above, with the exception as noted in item (a), have been satisfactorily completed.
For the foregoing reasons, the Commission has found that the public heal th, safety and interest require that this Order be effective immedi-ately.
I!!.
Copies of the following documents are available for inspection at the Commission's Public Document Room at 1717 H Street, H.W, Washington, D.C.
20555, and are being placed in the Commission's local public document roca in the Ida Ruoc Public Liorary, 310 Madison Street, Port Clinton, Ohio 43452:
(1) Office of Nuclear Reactor Regulation Status Report on Feedwater Transients in B&W Plants, 'pril 25, 1979.
2310 212 e
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7590-01 (2) Letters from Lowell E. Roe (TECO) to Harold Denton (NRR) dated April 27 and May 4,1979.
IV.
Acccrcir. gly, pursuant to the Atomic Energy Act of 1954, as amended, and the Commission's Rules and Regulations in 10 CFR Parts 2 and 50, IT IS HERE3Y CRDERED THAT:
(1) The licensees shall take the following actions with respect to Davis-Besse 1:
(a) Review all aspects of the safety grade auxiliary feedwater system to further upgrade components for added reliability and performance.
Present modifications will include the addition of dynamic braking on the auxiliary feedpump turbine speed changer and provision of means for control room veri-fication of the auxiliary feedwater flow to the steam generators.
This ceans of verification will be provided for one steem generator prior to startup from the present maintenance outage and for the other steam generator as soon as vendor-supplied equipment is available (estimated date is June 1,1979).
In addition, the licensees will review and verify the adequacy of the
- uxiliary feedwater system capac ;ty.
(b) Re.ise operating procedures as necessary to eliminate the option of using the Integrated Control System as a backup means for controlling the auxiliary feedwater system.
23l0 213
7593-01.
(c)
Implement a hard-wired control-grade reactor trip that would be actuated on loss of main feedwater and/or turdir.e trip.
(d) Complete analyses for potential small creaks a-c cevelc:
and irplement c;erating instructions to cefir.e cre-ator action.
(e) All licensed reactor operators anc se1ior reac:;r cperators will have completed the Three '<ile Island Unit No. 2 sirul a:Or training at B&W.
(f) Submit a reevaluation of the TECO anz. lysis of tne need for automatic or acministrative control of steam generator level setpoints curing auxiliary fe?csater syster operction previcusly submitted by TECO letter dated December 22, 1978, in licht cf the Three Mile Island No. 2 incident.
(g) Submit a review of the previous TEC0 evaluation of the September 24 1977 event involving equipment problems and depressurization of the primary system at Davis-Besse 1 in ligat of the Three Mile Island Unit No. 2 incident.
(2) The licensees shall maintain Davis-Besse 1 ir a shutcown ccncition until items (a) through (g) in paragrapn (1), except as nc ec in itam (a), above are satisfactorily completed.
Satisfactory cca;1eti:n w'll req; ire confir-nation by tne Director, Of fice of Nuclear Reactor Re;u'a-icn, t a: the 2310 214 o
. actions specified have been taken, the specified analyses are acceptable, and the specified implementing procedures are appropriate.
(3) The licensees shall as promptly as practicaole also accomplish the long-term modifications set forth in Secticn :: of tr.is Order.
V.
Within twenty (20) days of the date of this Order, tne licensees or any person whose interest may be affected by this Order may re;uest a hearing with respect to this Order. Any such request shall not stay tne immeciate effectiveness of this Order.
FCR THE NUCLEAR REGULATC.Y CC".*11SSION Samuel J. Chilk Secretary of the Commission Da ted at Washington, D.C.,
this day of May 1979.
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,,.s..;..,..y u.: r-t. o Docket No. 50-346 Liceuse Ko. NTF-3 Serial Kc. 497
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M. Harold R. Danton, Director C" ice of Nucicar "cactor Regulation U.S. Suc1 car T.ogula tory Con-ission Wenhington, D.C.
20555
Dear Mr. Deuton:
In ycur teeting of /.pril 24, 1979 with represen:atives of E2hcoch 5 Wilco:c and four lice sees, including Toledo Edison, who have B5W nuc1 car stca:
supply systc=s, a nt=ber of concerns were c pressed by you sad your sgaff regarding cert:in features of the E5W plants.
Tnese concerns were further detsi3cd in 3 our h*RR Status Report on Feedvater Tecnsients in '5V Plsnts 8
g of April 25, 1979.
In this report, on page 1-7, certain suggested n cps t/
vere outlined which, if takeu, would provide assurs.nce to you. hat the I4W plants could continue to operate without undue risk.
l o
Unile ve feel that a nunber of design features already incorper::cd in the i
D.tvis-Sesse Unit.1 fully rect or exceed the criteris you are fr. questing h
sed that Davis-Besse cen be operated without undue risk, vc cre prepcsing f't the following actions:
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A.
Au.iliary Peedeste_r S.ystem Relia _bility an d.P.e. r.f.o.r...an c e 2310 218 h"i o
The atu:iliary feedvater sys:c= for the Davis-Ecsse Unit 1 is =
l reliable full safety grade syste= vich redundancy for =eerfur, the F.
single failure <.:riteria.
"Ihe principal features are detat. led in h-'
Tabic 2.1 of your report.
f Mi
,'p Mc, however, vill continue to reviev all aspects of this system to further upgrade components f or added reliability and peric re:nce.
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Coe such ite: 12 an installation of dyna =ie brakiur, on the: s tra li cr:-
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fccd punp turbine speed changer to f urther einici::e icvel fiocruninn E.
f In the stea: ;;cncra tor when on au>:lliar,c feed, 3.
n es:rcred Cont rol S.ste (ICS) In fl uence or. Ate:i l i a r.- reedvat:r 'e.:rol l.
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.he Davis-E.c.:sc cu::iliary fcedvater control systa: iv : full s.- f.- t y
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grr.x sys:e cc:cletely independent of ICS.
Tne euxilicry f eeduc.: er
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- : t i -: : co: :rol is es;ahle of bein; r.titched to ICS for si hseL ur - e
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Anticinatory Ara = of Reactor j
a
/tddition of a hard vired control grade reactor trip on loss of cain E
feedvater or turbine trip.
U
,D.
S:all lireak Analysis j
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' hark with Ef.V to corp 3 cte the analyses for potential s=.tl1 breaks and d::vclop :nd i=plecont acy necessary operating, procedures to define operator action.
E.
0::ratinr Proe._ durn and Curator Trainir.3
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All proct:'urec needed to be developed or r:odified by actiens A th.m D i131 be ce=picted and training of the cperators in the proceslures
. vill be done.
All licensed shif t opera tors will i ave received.MW ni=ulatcr trening on the !?.I-2 incident.
/dl rf the prcpeced actions outiiued in A thru D above vould be taken prior to s t:rt-up frc= the current rtaicten nce cutsge.
Toledo Edison v-ill centinue ef forts to pros-ide additional it prevere::ts related to A thru D as follows:
y.
A.
Centinue to revice perfer=ance of t.he sys:en fcr essurance of reliability l.
and perfcr: ance.
f-B.
.The failure code and effects cr.alysis of ICS is ur. der voy with priority by 24W and vill be sub itted as soon as possible.
C.
Tne reactor trips vill be revised to safety grade as far as possible.
D.
Centinuing attentio= will be (Lven to trannient analysir, and procedures f er canagercat of s=al) breaks.
E.
Continue cpera tor training and tretininn as a part of our ouncing p r o g re:.- to continue to assure the hilgh state of readiness of our operating r caf f.
Uc are confident that these actions on our cart vill satisfy yent conectr.s p
and ptr. vide additional and full assurance for public safety,
[b r
Yours ecry t ruly,
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Docket No. 50-346 A
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lir. E;rold R. Denton, Director h
Of fice of Nucicar hactor Regulcticn U.S. Fu:Ie:r Ogulstory Coccission lia r.hin s t e n, D. C.
20555 U.
D# ar "r. Dm.::ca:
p' 24, 1979 with representatives of Babcock 6 Vilcox p
In your meeting of April includice Toledo Edison, who have B&W cuclear ates:
y e nu ber of concerns were expresr.ed by you sed your etaff
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ccd four Ifernnecc.,
cupply 3 p tenn, These concerr.s vere further reprding certain festures of the SW plants.
det.:.11ed in your hRR St.tus Eepert on feeduater *fransients in EW I'1 Ants
- i of gril 25. 1979. In this report, on page 1-7, certain suggested steps j
were outlined which, if taken, vould provide assurance to you that the p!
i' E',W planta could centinue to operate without (mdue risk.
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kalle va feel that a n: 6er of de:ign f eueres aircady incor7 orated in the Davis-Passe Unit I fully neet or exceed the criteria you are requesting i
and ther Davis-Ecrue can be operated without (mduc risk, we ate propesing
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p the folle:.rf ng actions:
- i Aux 1115,ry_F,eydvater Sr9 tem Reliability and_,P,er forpnce M
A.
The auxiliary feedwater syntes for the Davis-Besse Unit 1 ir.
(:'
reliable full safety Ersde system with redundancy for necting the h
single failure criteris. The principal features are det=iled in y
tt Tahic 2.1 of your report,
..bug Wa, however, vill conticcs to review all aspects of this syste.'s to Q
for added reliability and perforcance.
further upgrado cc:rponentti J
One nuch iten is an installation of dynamic braking on the aux.111ary i
feed pump turbine speed changer to further minicize level fluctuation j
in the oceam generator when on auxiliarf feed.
H D.
Integrated Control Sya,te (ICS) Influence oc Auxiliary, Feedvater Control h
The Devic-Easce auxiliary feedvater control system to a full safety Toc auxiliary feedveter j
grade syst+c cc=pletely independent of ICS.
r.-aster control is capable of being evitched to ICS for a beckup neans
{}!
of control, but this option is to be rescVed itsediately by administra-p g.
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tive procedures. (
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4
!M' 2310 220
- E n
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.. ' Mr. Escold R. Denton, Director r 1 27, 1979
)
_Antiy1Lat,ory Sc ram _o f_R,eac tor C.
l b
Addition of a hard wired control grade reactor trip on loss of main 4
D.
Sena11 Break Analysis k
3 Work with B&W to complete the analyaca for potential small breaks and I
develop and implement any necesssry operating procedures to define l
operator action.
t E.
Operating Procedures and Deerator Traininc
(
b All procedures needed to be developed or modified by actions A thru D
{
will be completed and traininr. of the operators in the procedurce will be done.
All licensed shift opzrators will have received B5W simulator trining on the 7211-2 incident.
I i A.l1 of the proposed actions outlined in A thru D above would be taken prior t
to start-up from the current maintenance outage.
g n
Tolede Edison will continue efforts to provide additional improvements
[
relat_d to A thru D as follows:
l A.
Continue to review perfor=ance of the system for assurance of reliability and performance.
p w
B.
The failure mode and effects analysis of ICS is under way with priority by B&W and Vill be subnitted as soon as possible.
L C.
The reactor trips will be revised to safety grade as far as possible.
D.
Continuing attention will be given to transient analyais and procedures for management of anall breaka.
[
E.
Continue operator training and retraining as a part of our ongoing program to continue to aasure the high state of readiness of c tr operating staff, Y.
We are confident that these actions on our part will satisfy your concerns E
and provide additional and full assurance for public safety.
I:
ki Yours very truly, f
f k
2310 22 I
imall E. Roe Vice President t
Facilities Develop:sent h
The Toledo Edison Company j
i" r
I-tu
7590-01 UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION In the Matter of
)
)
FLORIDA POWER CORPORATION, ET AL )
Occket No. 50-302
)
Crystal River Unit No. 3
)
Nuclear Generating Plant
)
CRDER I.
Fiorica Power Corporation (FPC or the licensee) and eleven otner cc-caners are the hoicers of Facility Cperating License No. DPR-72 which autnorizes the operation of the nuclear power reactor known as Crystal River Unit No. 3 Nuclear
-enerating Plant (the facility or Crystal River Unit 3), et steacy state p wer levels not in excess of 2452 megawatts thermal (rated pcaer).
The facility is a Sancock & Wilcox (B&W) designed pressurized water reactor (P'JR) located at the licensees' site in Citrus County, Fiorica.
II.
In the course of its evaluation to date of the accident at the Three Mile Island Unit No. 2 f acility, which utilizes a B&W designed PWR, the Nuclear Regulatory Commission staff has ascertained that B&W designed reactors appear to be unusuali sensitive to certain off-normal transient conditions originating in the secondar;.
system.
The features of the B&W design that contribute to this sensitivity are:
(1) design of the steam generators to operate with 'relatively small liquid volume in the secondary side; (2) the lack of direct initiation of reactor trip upon thc 2310 222 e
e
7590-01 l
L t occurrence of off-normal conditions in the feedwater system; (3) reliance on an integrated control system (ICS) to automatically regulate feedwater
. b ficw; (4) actuation before reactor trip of a pilot-operated relief valve on the primary system pressurizer (which, if the valve sticks open, can aggravate the event); and (5) a low steara generator elevation (relative to tne reactor vessel) whicn provices a smaller driving nead for natural circu-l ati en.
Eecause of these features, B&W cesicned reactors place core reliance on tne reliacility and performance characteristics of the auxiliary feecxater system, the integrated control system, anc the emergency core cooling system (ECCS) performance to recover from frequent anticipatec transients, such as loss of o
of fsite peser and loss of normal feednater, tr.an do other PWR cesigns.
- Tnis, in turn, places a large curcen en tne plant operators in tre ever,t of of f-
- {
normal system behavior during such anticipated transients.
As a result of a preliminary review of the Three Mile Island Unit No. 2 accident
~
chronology, the NRC staff initially identified several human errors that occurrec All holders during the accident and contributed significantly to its severity.
of operating licenses were subsequently instructec to take a numoer of immediate actions to avoid repetition of these errors, in accordance witn oulletins issued oy the Commission's Of fice of Inspection and Enforcement (IE).
In 3dci tior the NRC staff Degan an immeciate reevaluation of the aesign features of B5W 2310 223 e
7590-01,
reactors to determine whether additional safety corrections or improvements were necessary with respect to these reactors.
This evaluation involved numerous meetings with B&W and certain of the affected licensees.
Tne evaluation identified design features as discussed above which indicated that E'u cesignec reacters are unusually sensitive tc certain eff-ncrmal transier.: cone"tions originating in the secondary system.
As a resul t, an aaditicnal bulletin was issued by IE which instructed hoicers of operating licerses for S&W desigr.ec reactors to take further acticns, inclucing immedicte changes to decrease the reactor high pressure trip point and increase the pressuricer pilot-operated relief valve setting.
Also, as a result of this evaluation, tne NRC staf f identified certain otner safety ccncerns that warranted additional short-ter.T design and procedural ch'anges at operating facilities having E&W designed reactors.
These were icentified as items (a) through (e) on page 1-7 of the Of fice of Nuclear Feactor Regulation Status Repor-to the Commission of April 25, 1979.
Af ter a series of discussions between the NRC staff and the licensee concerning possible design modifications and changes in operating procedures, the licensee agreed in a letter dated May 1,1979, to perform promptly the following actions:
2310 224 e
S
7590-01 (a)
Upgrade the timeliness and reliability of delivery from the Emergency Feecwater System by carrying out actions as identified in Enclosure 1 of the licensee's letter of May 1,1979.
(c)
Develop and implement cperating prccecures fcr initiatinc and controlling e ergency feed ater incependent of Inte-grated Control System control.
(c' Implement a harc-wirec control-grace reactor trip tnat would be actuat>J co loss of main feeceater and/or turbine trip.
(d)
Cor?lete analyses for potential small breaks and cevelop and implement operating instructions to define operator action.
(e)
All licensed reactor operators and senior reactor operators will I
have completed the Three Mile Island Unit No. 2 (TMI-2) simulator training at B&'n'.
In its letter the licensee also stated that the facility is shut down and would remain shut down until (a) through (e) above are completed.
In addition to these modifications to be implemented promptly, the licensee has alsc proposed to carry out certain additional iong-term modifications to further enhance the capability and reliability of the reactor to respond 'to various transient events.
These are:
2310 225
7590-01
-5 The licensee will make modifications to provide. verification in the control room of 'mergency feedwater flow to each steam generator.
The licensee will submit a failuca mode and ef fects analysis of the Integrated Control System to the NRC staf f as soon as prac-ticable.
The licensee stated that this analysis is now uncerway with high priori ty by B&W.
The reactor trip following loss of main feedwater and/or trip of the turbine to be installed promptly pursuant to this Order will thereaf ter be upgradeo so that the componcats are safety grade.
The licensee will submit this design to the NRC staff for review.
The licensee will continue reactor operator training and drilling of response procecures to assure a high state of preparedness.
The Commission has concluded that the prompt actions set forth as (a) through (e) above are necessary to provide added reliability to the reactor system to respond safely to feedwater transients and should be confirmed by a Commission
~
order.
The Commission finds that operation of the facility should not be resumed until the actions described in paragraphs (a) though (e) above have been satisf actorily c ompl e ted.
2310 226 e
7590-01 For the foregoing reasons, the Commission has found that the public he al th, safety and interest require that this Order be effective immeciately.
III.
Copies of the fellcwing cocuments are available for inspecticr. at the Commission's Puolic Document Room at 1717 H Street, N.W., Washincter.,
D.C.
20555, and are being placed in the Commission's local public cocurar.: room in the Crystal River Public Library, Crystal River, Florida, 32529:
(1)
Office of Nuclear Reactor Regulation Status Report on Feecwater Transients in B&W Plants, April 25,.1979.
(2) letter from B. L. Griffin (FPC) to Harold Denton (NRR) dated t'ay 1, 1979.
IV.
Accorcingly, pursuant to the' Atomic Energy Act of 1954, as amended, and the Commission's Rules and Regulations in 10 CFR Parts 2 and 50, IT IS HEREBY ORDERED THAT:
(i)
The licensee shall take the following actions with respect to Crystal River Unit 3:
(a)
Upgrace the timeliness and reliability of celivery from the Emergency Feecwater System by carrying cut actions as icentified in Er. closure 1 of the licensee's letter of May 1,1979.
2310 227 6
7590-01 (b) Develop and implement operating procedures for initiating and controlling emergency feedwater independent of Inte-grated Control System control.
(c)
Implement a hard-wired control-grade reactor trip that would be actuated on loss of main feecwater and/or turbine trip.
(c) Complete analyses for potential small creaks anc cevelop and implement operating instructions to cefine operator action.
(e) All licensed reactor cperators and senior reactor operators will have completed the TMI-2 simulator training at E&W.
(2) Tne licensee shall maintain Crystal River Unit 3 in a shutdcwn concition (the f acility was shut down on April 23, 1979) until items (a) through (e) in paragraph (1) above are satisfactorily completed.
Satisf actory completion will require confirmation by the Director, Office of Nuclear Reactor Regulation, that the actions specified have been taken, the specified analyses are acceptable, and the specified implementing procedures are appropriate.
(3) The licensee shall as promptly as practicable also accomplish the long-term modifications set forth in Section II of this Order.
2310 228 e
7590-01.
t i
Y.
Within twenty (20) days of the date of this Order, the licensees or any person whose interest may be affected by this Order may request a hearing with respect to this Order.
Any such request shall not stay the ir.metiate effectiveness of this Order.
FOR THE tiUCLEAR REG'JLATORY COMMISS10t,'
i 4
Samuel J. Chilk Secretary of tne Commission 4
Dated a t 'a'a shi ng ton, D.C.
this cav of Mav 1979.
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2310 229 B
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. Direct.er Cf fice of i;uc k neactor Regulation U.S. Tiuclear r.:.. a r cry Cor.:ci s s i cr Washingtcr., D~
a:55 Sub;ect:
Cryst;l Ri ver Uni t 3 Docket 7,c. 5C-302 Operating License CPR-72
Dear P. Centon:
- n res;onse t: Staff safety cor.cerr.s icentified as items a. through e.
- n peue 1-7 of ae 0"RR Status Report to the Corrission cf April 25, c.;er Corporation proposes to inplement the following iW, iorica o
interna actions until further analysis of these concerns can be cor.p l e t ed :
(a)
Upgrade of the timeliness and reliability of delivery from the Emergency Feedwater System by carrying out items 1 through 9 identified in Enclosure 1.
(b)
We have developed and implemented operating procedures for initiating and controlling emergency feedwater independent of ICS control..
Implement a hard-wired control-grade reactor trip on loss of (c) main feedwater or turcine trip.
(d)
Complete analyses for potential small breaks and develop and implement operating instructions to define c; erator action.
(e)
All Cor. trol Room operators have completec TP.1-2 training on the 8F/.. simulator.
Crystal River Unit 3 is shutdown for maintenance and refueline and Florica Power Corporation has connitted in our April 27, 1979, letter to you to resolve and inplement items a. through e. pricr to startup t;hich is currently scheduled for June 1, 1979.
2310 230 I,
b N b2Ih.. I. [ [,
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Harold R. Denton Page 2 May 1, 1979 FPC further commits to the following ongoing /long term actions for improvement 'and' assuring safety at Crystal River Unit 3:
The failure mode and effects analysis of ICS is undernay (a) with high priority by S&W and will be submitted as soon as practicable.
Upon completion of a detailec design and supporting analy-(b) sis, the hard-wired trip will be revised to a safety grade syster.
Modifications will be race te provide verificaticr, in the (c) centrol room of EF'n fic t: each stea genera:cr.
k more corplete description cf the small breaks transient
( d ',
analyses is provided in Enclosure 2, entitiec " Guidelines f c-tne Develeprent of Operational Fracedures for Safe of Small Breaks in the Reactor Coclant Syster Management Pressure Soundary."
We will continue operator training and drilling of response
~
(e) procedures as a part of our ongoing program to assure the high state of raadiness and safe operaticn at CR3.
We are confident that this action will meet your Staff concerns and pre-vide additional assurance of the health and safety of the public.
Very truly yours,
. ) i,i,,l-
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2310 231 4
S. L. Griffin PYSekcS01 D65 Enclosures e
SI ATE OF FLORIDA COUtlTY OF PINELLAS B.L. Griffin states that he is the Senior Vice F resident, Engineering and Construction, Florida Power Corporation; that he is authorized on the part of said cc:npany to sign and file with the Nuclear Regulatcry Cor: mission the infor: nation attached hereto; and that a ll such statements riade and r.atters set forth therein are true and correct to the best of his knowledge, information and belief.
j,
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, i ii. L. 3r:f f;n Subscribec and sworn to before im, a Notary Public in anc for the State and County above named, this 1st da. of May,1979, JE 2
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b Notary Public ~
Notary Public, Stat e of Florida at Large, 2310 232 i'y Coanission Expires:
July 25, 1980 (Notary 1 D12) e O
e
D :tOSURi (!)
May 1. Ioli>
AUXIL! ANY TECDh'ATLR SYSTU1 UPGRADL Review crocedures, revise as necessary and conduct training to en-sure timely and proper starting of motor driven emergency feedwater (EFL') pum;; frcm engineered safeguards bus A upon less of offsite p oi.e r.
To assure that EFW will be aligned in a timely manner to iniect on 2.
ali EFW cemanc even;s when in the surveillance test mode, proce-dures will be 1c.olemented, and training conducted to provide an crera;cr at the necessary valves in com ;nication with the control rc:r caring the surveillance rede to carry cat the sa l ve ali gun'en'.
.:harc n scen EFh de.:and event;.
2.
r erye-cy te ec.;-1:- Oy:, ass valves are normaliy i the O_ ;
- Ci-i rccea.; e; '.e ve See-ceveicrec and i r; ie. e r. c tc rec: i re t c-tF'..
d '. c r tc 'cee ccr'rc' C these.alves u; c' failu-e c' the 1.- S t e: r 9er.e 3;;- leve; CCntrol.
If tr.e ICS.e'.ei cc-teci does 1
f ail :ne crerator will close the bypass valves.
Inose.;Ives the in he EFL syst e nc; lock ed in positior, a re ver-fi ed to be in cr:. -,cs. tion c, a cai.'.
casis.
Training w l' be concoci en t'en rev';ed _' ::edures prior tc Jane 1,19 79.
The EF,.' p res will be verifiec operable in accorcance with the LR#3 Techni cal Scecifications and Surveillance Procedures.
Re.;e, m c. c.. s e, a s r.e c e s s a ry, the proce ' ares and rairing for
- r s'dir-alt erna;; scurces cf water to tne suc
- 1cn cf tne EF.
pu cs.
6.
Re cve the interlock 5.hich prevents,the turbine-criven energency f eec..ater cump operation wlen the rotor cr; ven en&rcency feed..:Ler pump is rmini ng.
7.
In event emargency feecwater is necessary and offsite power is available, an auto start signal will be provided to the motor driven emergency feedwater pump.
as necessary, will be conoucted to 8.
Desi gn revi ew and modi fi cat. ion,.
- rovice contrcl room annunciation for auto start conditions 0;. the EFW system.
9.
Verification ha: been cace that the air operated level control valves (a) fail to the 507,open position upon icss c' pcwer to the electrical /pressere converter, and (b) fail to the a; positico a
upon less of irst u::.ent air and electrical pc..er to r.e air !cci.
At ful' load tnese valves are in.tne full (1007d op n positions anc at ic.. pc. er levels (below 10) they are partially ; pen centroliing f i c.s.
I' these vaives were to f ail closed, feecua 'r flow '::uld be centrolled using the EFW bypass.-alves as cescribe ir, iten.~
acove.
W g b 2310 233 s.-.
m etcso n us:
O e
ENCLOSURE '2) fiay 1. P379 GUIDELINES FCR TliE DEVELOPMENT OF OPERATIONAL PROCEDURES FOR SA E PANAGEMENT OF SMALL BREAKS IN T' E REACTOR COOLANT SYSTEM PRESSURE BOUNDARY Operational guidelines will be prepared for the safe handling of small breaks as an extension of and addition to previously issued guidelines and IE Bulletin 79-05A. These guidelines will include provisions for operator recognition of small breaks and discriminatien of other acci-dents which might procuce similar symptoms.
The guidelines will include exp'ected system response insof ar as required to assure ef fective operator understanding and action.
The guidelines will include necessary precautions and will describe those actions which the operator must take to assure safe management and mitigation of small breal events, inclucing natural circulation cooling w: ere it is predict-ed to occur in the course of the acticent.
These guicelines will specifically cover cases in which RCS stabill:a-both tion will occ.ur with a partially filled reactor coolant system for the case with' the reactor coolant pumps on and the reactor coolant pumps off. Delay in the initiation of auxiliary feedwater up to 20 minutes will be consics ed.
System conditions covered will assume availability of ECCS systems at full design flow in the event that auxiliary feed-water is not available or with single f ailure in the ECCS systens in the event that auxiliary feedwater is available.
The guidelines will be based on existing analyses and by specific addi-tional computer calculations.
These calculations will ce performed to define system response to restart of reactor cociant pumps in a partial-ly filled system and response of the partially filled system to restart of auxiliary feedwater.
These guidelines will be devloped by BSW and reviewed by the NRC staff in time f or implementation of the corresponding procedures cy Florida Power Corpo-ation on or before startup.
e a
2310 234 PYEetcS01(D65)
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UNITED STATES
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y NUCLEAR REGULATORY COMMisslON o,
yh WASHINGTON, D. C. 20555 8 M4h S
' 0, h3y l
1 FEB 11 1978
.. m
.S. Hw & nv y a
Docket No. 50-346 C L ouven
- h. J'e4whcc 1 LICENSEE: Toledo Edison Company s
FACILITY:
is-Besse uclear Power Station, Unit No. 1 (08-1) d 6.k
SUBJECT:
SUMMARY
OF MEETING ON NATURAL CIRCULATION TEST - (DB-1)
On February 7,1978 representatives from the Toledo Edison Company (TECO), the Babcock & Wilcox Company and the Bechtel Corporation met with the NRC staff to present their bases for not needing to conduct
.,~
a natural circulation test for DB-1.
A list of attendees is provided e
in Enclosure 1.
l At 2243 hours0.026 days <br />0.623 hours <br />0.00371 weeks <br />8.534615e-4 months <br /> on November 29, 1977, DB-1 experienced a transient
' (temporary loss of 13.8 VJ power) which tripped all four reactor i
i coolant pumps, and for approximately 15 minutes until 2258 hours0.0261 days <br />0.627 hours <br />0.00373 weeks <br />8.59169e-4 months <br />, reactor decay heat was removed by natural circulation.
During the I (
15 minute period the data-recording reactimeter was in operation and
\\.
TECO analyzed the data to see if natural circulation could be justified during the loss of station power transient.
TECO concluded from their analysis of the data that the transient did not satisfy the NRC test requirements for a natural circulation test.
Because of (1) the imbalance of the once-through-steam generators (OTSG) during the transient, (2) the lack of data for the loop 1 hot leg temperature, and (3) the non-equilibrium state of the NSSS during the transient; TECO cou,ld not analyze and qualify the transient as a satisfactory natural circulation test.
The NRC staff concurred with TECO that the loss of site power transient did not confirm a steady-state natural circulation flow rate required for the natural circulation test.
TECO then reiterated their previous position that the elevated position of OTSG's for 08-1 would increase the natural circulation flow of 0B-1 above that observed by test for Oconee No.1.
A summary of j
TECO's position is provided in Enclosure 2.
i TECO stated that the test procedures for conducting a steady-state natural circulation test would include 14 hours1.62037e-4 days <br />0.00389 hours <br />2.314815e-5 weeks <br />5.327e-6 months <br /> at 5% of full power to reach stable Xenon conditions and an additional 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> for both i
phases of the test.
Similar tests conducted in the past have required about three days.
TECO stated that the present coal supplies available for electric j
power generation in the state of Ohio have reached a critical point i
and 08-1 cannot be off-line and at reduc 5d power for the time required 2310 235 i
e
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c, to run the natural circulation test without impacting load requirements for their grid system and increasing the use of rapidly diminishing coal supplies.
The NRC staff stated that the question of requiring 08-1 to run a natural circulation test had been considered prior to issuance of the Operating License for DB-1 when the NRC staff had concluded that DB-1 was not considered as a similar plant to the prototypic Oconee1 since DB-1 was the first B&W 177 NSSS to use the elevated 0TSG's.
Both the NRC staff and TECO ind*cated they would be in contact with each other in the near future regarding these matters.
r M(di Leo'n Engl e,'q/-
Project Manager
?
l Light Water Reactors Branch No. 1
}
Division of Project Management
Enclosures:
1.
Attendance List i
2.
Summary of TECO j
f.
Technical Position 2310 236 i
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ENCLOSURE 1 ATTENDANCE LIST
-i FOR MEETING HELD ON FEBRUARY 7,1978
~
WITH THE TOLEDO EDISON COMPANY DAVIS BESSE, UNIT N0. 1 DOCKET NO. 50-346 Nuclear Reculatory Comission B. Clayton L. Engle C. Graves J. Mazetis P. O'Reilly D. Riehm
~
Toledo Ediosn Company C. Domeck F. Miller R. Sund 2310 237 Babcock & Wilcox Company i.
R. Davis J. Lauer C. Tally Bechtel Corporation D. Dismokes i
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ENCLOSURE 2 TO CUNllitCT A NA1PHA1. C1948.lll.ATION FEB 14137 7F.37 - DM'I n-hl ssr, UN 1'I 1
[b? !
,t t
lhl The4P which resul ts in nnturn) cJrenIntlun flow enn he exps enard by tiec f oi equntton:
Wir.)
f t.( pim)- 17 f t.[p f r'i)- 1.hple D{'
D D
r fd
- 'f" h.'.
O P -I Lp = Lepe
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" bb[
J 2
where O P - dif ferent int p cornire nvn l in t.le for uninnal cfsenintinn flow I.c vert J eal di at nnce f i nm I.o t t ori. ni enr e to botrom of t er'ipern t ni c t r nnni t S een r.onc in nfenm prncrninr.
'l dennity of told leg water pc
=
I ft dept h of temprenture trannitInn ron" fu ascom y. cur s at os' s
plm = log menn dennity of wntc 12 ft - active core J enp.t h (t rona f t f on rone)
Lh = vertlen) distnoce f om top of cais e fn t e.p t cinpes nt no e i r nonii loo r.on c,13: ntenm generator ph = dennit y of hot ]ct, unter 1,c (Oconee) 4.5 ft 1.h ( oc..n..r )
14 f
Le (nnvin-Besse) 68.(i ft
=
i t. (li.v i n ncn.ic)
.'. /. t.
It pc 47.7 lla./ I t
=
ph 4(i. 7 lb / I Compa ri ng, Oconec 1 an.I Davin neuse I:
Increnned d r - (21
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2.'l. fi, f. 'i)
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144 v
0.25 lb/In!
e The only difference bet ween oconce I nud pnvin-D.'noc 1 i t: In IIn-rnined uncom gencrntorn, which only tend to inercnnc 6P Jn 15a v i s-De a ne 1 nnd henc e i nri en ne flow (Q).
For t.he system curve t hat app] fen t o I.nr h 0conce 1 nnd linv in-nenne !
T
'MVIS BES5E-1 s
(_ i CO OcoN EE Cte,.
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Graph 1 also shows calculated flow for Davis-Besse 1.
November 29, 1977 transient data shows that during 15 minutes that reactor coolant pumps were idle, heat was being removed (loop Tb and TC decreasing).
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P.O. Box 1260. Lynchburg. Va. 24' y.
Telephone: (703) 384 5111
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July 16, 1975 N,
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~ SOM #055-620-001!.
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IEECEgygg EO1 Mr. J. G. Evans, Station Superintendent fg,.9Q Davis-Besse Nuclear Power Station 5501 North State Route #2 Oak Harbor, Ohio h3hh9 Sub.j ec t :
Justification for Deletien of the NSSS Natural Circulation Test
Dear Jack:
n v
Our Engineering Depart =ent has completed an in depth analysis of the USSS natural circulation characteristics of Davis-Besse Unit I.
The analysis was based on the Duke Power Company Ocence I natural circulation tnst re-suits and an analytical extrapolation of results by comparison of Oconee I and Davis-Besse Unit I designs. As a result of this analysis, it was con-cluded that, Davis-Besse Unit I vill exhibit more natural circulation flov than.0conee I due to the elevated position of the steam generators with re-spect to the reactor vessel.
In lieu of i.he Oconee I natural circulation test results and the analysis by our Engineering Department the following is recommended:
l.
The natural circulation steam generator water level setpoint, Integrated Control System FW 20.4 and FW 21.9, should be reduced from 955 to 50% on the operating range instrumentation.
505 is the setpoint used in the
~
Oconee I natural circulation test and the engineering analysis.
2.
The natural circulation test at Davis-Besse Unit I need not be " conducted
. per Regulatory Guide 1.68, Appendix A, Part D.la (November,1973) which states as follevs:
" Natural circulation tests to confirm sufficient cooling capacity.
Com-(
parison of adequate flow data vith the performance of previously tested plants of like design may be substituted for this test."
Ir order to support item 2 above, the following information should be useful in d-leting the requirement fbr a natural circulation test.
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"ggspj))' 17 TP 800.04 h!g
}f,1.n -77 Nb
~LAH D4 Vis 3cLSc Nuc1 Car Power Station wHS
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Unit No. 1
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RECEIVED gr i r s_' P
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-aw' Y e.u o2.-i.
U:. l LuD~~
~~ ]_ gf/gg.:ocedure TP 800.04 9g0 021977 1.^
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POWER ENG
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[~Na ;' tra1 Circulation Test
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ASD 1
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($:cio6 Head Date Seco:=c t
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Date
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A bnagt-of Quality Assur v e
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Approved by
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Station "6peYiatendent
' Da te Revision SPJi QA Stv, Supt.
No.
Re,c oc=enda tion Date Approved Date Approved Dat
/shf/77 l'I 5 Q t} W
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Please Retura Pievious Revision to the DavisV Betse Office' Supervisor, Stop.2103 /
TEST PEP.FORMNCE Test Co:spleted Test Leader Date i
./
Recomended by-
\\.
Scction Head Date j
Recor.cnded by 9
51.1 Chairman Date f
~
~
2 3.1 0 2.4 2.
s e
1 TP 800.04.0
[
1.
PURP E 6"
The purpose of this precedure is to verify that on loss of all forced reactor fj- {
coolant ficw, natural circulation will begin to provide adequate core cooling V i for all possible levels of decay heat generation. The procedure provides two e
methods of measuring primary flow rate under natural circulation conditions.
~
i j
We test is perfonned in tw phases.
Since under natural circulation conditic j
Tc will change significantly frcm the value at which p:rer range instn:mntat-J was calibrated, it will be necessary to measure the effect of a change in Tc (
indicated neutron flux. his is acccmplished in Phase I.
Phase,II is the natural circulaticn test itself. With reactor pcwer at 2 - 4% FF/ Auxiliary Fecdaater flcw is established to the OPSG's. While naintaining the reactor criticid., the oprating PC pumps are trippd. h*nen steady state conditions are established, primary flcu will be measured by calculation using reactor A T and by measurcment of loop transit tinn.
2.
D'UIP:eir IE.EDED 2.1 Reactimeter 2.2 Brush Recorder (6 Channel)
Eqaip. No.
2.3 Digital Volt:mter
~
Equip. No.
3.
REFERENCES 3.1 Davis-Eesse Final Safety Analysis Paport, Section 15.2.5.
3.2 Power Escalation Controlling Procedure, TP 800.00.
Reactor Protective Sys. tem Operating Procedure, SP 1105.02.
3.3 3.4 Physics Test dbnual (B&W) 'IU 000.23.
3.5 TECo Nuclear Quality Assurance bhnual.
s 3.6 AD 1801 series on the Conduct of the Preoperational & Startup Test Program.
3.7 Auxiliary Feedwater Systcm Operating Procedure, SP 1106.06.
3.8 Pcwer Operations, "PP 1102.04.
s
{'IS) 3.9 Davis-Besse Technical Specifications:
r;
{
TS 3.10.3 ' Special Test Exception - Reactor Coolant Icops s.
TS 3.4.1 Limiting Ccnditica for Operation - Rea.ctor Coolant Icops TS 3.2.5 Limiting Ccnditica for Operation - DNB Para:mters TS 3.1.1.4 LUniting Ccndition for Operation - Flinimt n '1bnperature for Criticality
l,Y. '
f 2
TP 800.04 1 M.
- (.
'Is 3.1.1.2 Limir.ing condition for Operation - Boron Dilution d'. '
'Is 3.7.1.3 Limi':ing Condition for Operation - Condensate Storage Tank e
3.10 SP 1103.15, Reactivity Balance.
I
'~
4:,
l i
3.11 ST 5030.11, RPS PcW r Range Calibration l
y-. - -
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3.12 TP 800.05, Reactivity Coefficients at Pwer 3.13 IC 2000.03, SettSg RPS O m p er Trip Bistable Setpoints.
~
3.14 ST 5030.02, RPS Ibnthly Check 4.
TIME & PERSOR3'L ILTUII"D__
l-4.1 Each Phase of tMs test will require approximately 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> to ccuplete.
Note that this d;2s not include the time required to establish the required plant c. xlitions, notably Xenon conditions, which will require l
an the order of I hours at w5% FP.
4.2 Personnel Recuir ;d, r.
Phase I
- Reac :or Operator at Dia:Tond Staticn
(.,
Reac mr Operator controlling feedwater Peac:ninter Operator Test L2ader Phase II - Reac:or CWator for Primary Plant Controls l
Reactor Operator at Diarond Statica Reactor Operator at feedwater controls Reactimeter Operator Brusa Recorder Oparator 1
Equip: ant Operator at Auxiliary Feed Pt:nps I&C Tc.:hniciais to jumper RPS, Measure NNI voltages, etc.
'Ibst Ieader Shift Foreman in Control Econ
/
5.
LD1ITATICNS & PRECAUTIES 5.1 When on natural circulation, manually trip the reactor and start one reactor coolant pump in each loop if any of the follwing limits are i
reached:
PAR M IER HIGi LDIIT ILW LIlm I
Indicated Reactor Pc e r (NIS,6,7, or 8) 5% FP*
(,
Reactor Ccolant Pressure (PRS RC2B2, PC' 2) 2300 PSIG 1990 PSK Pressurizer IcVel (LRS RCl4) 280 IN.
50 IN.
OISG Pressure (PI SPl2B or P'I SP 12A) 1020 PSIG
~
~
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r
,j 4
3 TP 800.04.1 s
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Any Incore T/c (7511 - T562) 650 F
~
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'Ih (TI RC3B1, RC 3Al) 600 F r
e, CST Invel (Tank in Service) 10 Pr.
7,, '
Corrected (See Section 7.1) u 5.2 hhen on natural circulation, manually trip the reactor and begin feedinc using a main feed pump it' at any time an auxiliary feed pt.7.p is lost or
, auxiliary feehater ficw.s lost for any other reason.
('IS) 5.3 Do not change reactor coolant Irron concentration during the time reactc coolant pumps are off.
(TS 3.1.1.2) 5.4 hhile on natural circulaticn, pressurizer spray will ba unavailable, therefore pressure control of the RCS will be slow at b2st. Tne. mans available are changing orSG 1evel using auxiliary feedwater, pressurizer heater control, letdoan & makeup, and (in an ar.ergency) electra atic relief. Fonitor temperature and pressurizer level trends and take action well in advance to minimize primary pressure excursions.
6.
PRERECUISITIES 6.1 Prerequisites for PHASE I.
6.1.1 The plant is at N 15% FP per PP 1102.04, Pcwer Op2 rations.
Verified Date 6.l.2 RCS boron concentration is + 30 ppnb of the value at which Phase II will be run, Verified Date*
Phase II will L2 conducted at 2 - 4% Fk', ap,3roximately equilibritr NOIE:
Xe, with GP. 6/7 60-80% hT).
"~
6.1.3 Feedwater darand stations FIC ICS 32B/32A are in IUWD. Fea3 water demnd is on Ioa Invel Limit.
Verified Date 1
6.1.4 Pressurizer Invel is a200"; Fakeup tank level is N 85",
and
~
there are 2 letacwn coolers in service.
Verified Date
/
(
6.1.5 The reactinuter is cet up to record data per Attac.Sent 1.
Verified Date 2310 245
=
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- *p e
S
' f.: "
4 TP 800.04.1 k,,'
6.1.6 Reactor Pwer Intalance is 0 + 0.5% FP.
q Verified Date
{., I 6.1.7 Grotip 6/7 position is 70-80% WD.
e.,
J-:
Verified Date C.
6.1.8 The NSS heat h11ance progra:n in the Plant Ccmpater has been lI
'l checked and is corputing core thermal pwer accurately.
Verified Date 6.1.9 'Ihe follwing parameters are on corputer trend reccrdrs on the operator's console.
IG R1 Flos (LDop 1)
F674 y
td M1 Floa (Laop 2)
F682 m.
tg J427 Verified Cate 6.2 Prerequisites for PHASE II t,
f j
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6.2.1 Tne plant is in pwer escalatica testing per TP 800.00, Pomr N.
Escalatica Scquence.
Verified Date 6.2.2 The follcwing testing has been carpleted at 15% FP:
{'
TP 800.05, Reactivity Ccefficients at Poer TP 800.08, ICS Tuning at Pomr TP 800.22, NSS Heat Balance TP 800.02, NI Calibration at Pcwer 1
Verified Date I
6.2.3 Pcactor Power is at 2-4% FP par PP 1102.04 with one main feedwate i
pump and 2 BC pumps in op2 ration. The Main ni Pump is en IIain Steam. Mini-feed to bcth S/G's is in operation.
Verified Date 6.2.4 Xenon concentration is approaching equilibritrn, such tlut calcu-lated reactivity change between the start of the test ard test carpletion will be less than.04%A k/k.
(Ref. Reactivity Balance Calculation, SP 1103.15)
\\-
Verified Date i'
2310 246
. -. -.. -. =
h-g
5 TP 800.04.1 f
6.2.5 Heat Balarce & NI Calibration (ST 5030.11, RPS Pcuer Pange g,
Calibration) luve been carpleted at 4% FP at the boron concen-g'g; tration at which the test will be run.
- b. -
Verified Date 3
6.J.6 A mininun of 47 incore thennoccuples (TE IM01 - TE 1 M14) are operable.
Verified Date 6.2.7 Moderator tenperature ccefficient has been reasural per TP 800.0 Reactivity Ccefficients at Pcuer, and is predicted to be no nere positive than 0.0 M k/k F at the poonr level and heron concen-tration at which this test will be run.
IUI'E: The core must inve been expended F4 EFPD tefore the temperature coefficient is negative.
6.2.8 Pressurizer IcVel is approxirately 150".
Pressurizer & Fakeup tank are within + 30 p;r.b of RCS concentration, @ere are 2 1
letdcron coolers in service.
Verified Date 6.2.9 We reactimeter & Brush Recorders are set up and calibrated to j
record data as specified in Attachmnt 3.
Verified
_Date 6.2.10 The Plant Ccnputer is set up to record data as specified in.
Verified Date I.
6.2.11 The ICS Configuration is as follows:
l
STATION STA'IUS HS ICS 1 Unit Master Track Turbine Tripped l
HIC ICS 13 S/G - RX Faster Hand HIC ICS 20 RX Demnd Track HC NI 44 Diamond Manual FIC ICS 32B FIC ICS 32A F/W Demand Hand (0%)
! f HIC ICS 30 A Tc' Hand (50%)
! (
HIC ICS 36A HIC ICS 36B Fain Feed Pumps, (one)
Auto FIC 1C3 353 l
FIC ICS 35A Thin FW Valves Auto FIC 1CS 338
~
e_'_.FTC T 'S '4 M...l...S /t' P: Val"-c g%
231.0'247 2
6
D '$.k h
p*
3 N
6 TP 800.04.1 bN e L-6.2.12 Tha RPS high flux. trip has been reset per IC 2000.03, Setting
- +
of RPS Overpcraer Trip Bistable Setroints, to 10% FP on all 4 RPS Channels.
('IS 3.10.3)
?. ?,
Verified Date 3
6.2.13 Both cond^.nsate storage tanks are filled with secondary makeup quality water to a level of '>30'.
Verified Date I
6.2.14 Tne Auxiliary Feedwater System, incitding both pgs, is operabir 1'
per SP 1106.06, Auxiliary Feedwater System. The Aux boiler is fi up to non.nl operating pressure.
Verified Date 6.2.15 Nron concentration in the ICS i; + 30 par.b of that when Phase I was c,ar.pleted. Batch calculations Fave been perfomed for additic 1
to the yakeup Tank.
Verified Date 6.2.16 'Ihe high flux trip portion of ST 5030.02, RPS Fcnthly Check, must (TS) b2 crr.ple'md within 12 hcurs of starting Phase II.
/
f Verified Date
\\
7.
P. E URE 7.1 PICSE I - Tha purpose of Phase I is to m2asure tha effect of re3uced tertparatures en indicated neutron flux. This data will be used to prcduce a co reccion to indicated pcwer which can be usei during the natural circulat. ion test.
7.1.1 Verify all prerequisites of 6.1 are canplete.
Verified Date 7.1.2 Obtain shift Foreman's parmission to begin this Phase.
Time Shift Forcman Verified Date 7.1.3 Adjust imbalance to 0 +.5% using the APSR's.
Verified Date 7.1.4 Shift Reactor Drand Station HIC ICS 20 & Diamond Station to
(
WAML. Paintain reactor pcuer at 15% controlling rods in s
Ka'UAL.
N Verified Date l.
7.1.5 Station c..Tothar reactor opirator at the feet. hater de: rand stations 1
to control feed.mter. Place pressurizer level centrol (IlC RC-14' l
- 1.. h. _'.
23'10 248
7 TP 800.04.0 g..
7.1.6 Start the trend recorders on the operator's console.
Jr l
Verified Date r.1 -
7.1.7 Begin recording data on the reactincter at 1 second intervals.
. Begin recording data on Attachnent 2 at 1 minute intervals.
Obtain NSS heat balance calculation fran plant cTpater.
(Pemrd a minimum of 5 minutes of steady state data).
Verified Date T:.
7.1.8 Using feedwater darand stations FIC ICS 32B/32A in HAND, SIciIX increase feedwater derard to lower Tavm5 F.
l NCTTE: This is an increase of onlyN4" SUR level and will reduct pressurizer level N 25".
- i...
j Do not m]ve control rods unless necessary to turn steadily in-creasing or decreasing pc7ser.
Verified Date 7.1.9 With Tav steady, balance feedwater ficw so that total feed ficw i-
.I is the same as remrded in step 7.1.6, 7.
It may at this point be necessary to adjust reactor power using control rcds in Fanual to achieve desired Tav & B1 ficw.
Verified Inte b~-
7.1.10 Wait for conditions to stabilize & recuest NSS heat balance frcm the plant carpater. Verify that Calculated Mith equals that determined in step 7.1.7
+ 20 Mith..
3:
Verified Date 7.1.11 Stop recording data on the reactimeter ("0" Switch), allcw the recorder to ccnplete its last record, then resu::n recording data.
Verified _
Date 7.1.12 Record 5 minutes of data en Attachmnt 2.
Verificd
_Date L -
7.1.13 Repeat steps 7.1.7 - 7.1.12 at w S incrorents until Tav = 559
+ 1 F.
b N.'
Verified Date 7.1.14 SIG1IX reduce ni dcmard until ni darand is on Icw level limit.
l Verified Date
. _ _. 2H0i2T9 4
e
e 8
TP 800.04.1 6.
ej:,4 7.1.15 Return the ICS to the lineup specified by the Shift Forman.
I' Verified Date s
n' s,
7-7 4
7.1.16 Stop reco -ding data on the reactimeter and data sheets.
s.
Verified Date p.
7.1.17 Inform the Shi_ft Fore:ran tPat this Phase is cmplete.
Verified Date 7.1.18 Delog rea::timeter data and determine correction factor for indic posar using the rinthcd on Attach cnt 8.
Verified Date d
~
7.2 PHASE II - Fatural Circulation Peasurcmant 7.2.1 Verify thn all prerequisites for PPase II, Section 6.2, hrie bc-t carpleted.
Verified Date a
7.2.2 Cbtain tM Shift Forman's permissica to b3 gin this Phase.
Time Shift Fore:ran Verified
'Date 7.2.3 Start taking data on the reactinnter at 0.1 second intervals.
. Start tha Erush Rccorder at 25mw' min.
Verified Date 7. 2.'4 Manually atart Auxiliary Feed Pumps 1-1 & 1-2 per SP 1106.06, Auxiliary Faedwater System, by opening MS 106 (HIS 106A) & PS 107 (HIS 107A) (CS717). With Fede Switches HIS 520B & HIS 521B (C57i In MMUAL, adjust speed to N2000 rpn.
1 NOTE:
Ee preparsi to add water to the MU tank as regaired.
Verified Date 7.2.5 Trip the SFRCS in the Icss of RC Pu:rp Fede by pressing & 3atchin HIS 4869E & HIS 4870S (C5721). This will align each Auxiliary Feed Pu:rp t'o its respective steam generator. Place press, level l
control in III'uND- (llc RC-14).
Verified Date 7.2.6 Sloaly & cvenly increase auxiliary fced pump speed using HIS 520.
and HIS 521A to raise steam generator level to a 100" on the SAJ Pange. The S/O Feedwater Valves will shut as level is increased This will cause a ccoldoen of about 9 F frcm 54PF and a corres-0 ponding 45" drop in pressurizer level which is desired. Insure 1
that mini-feed is in operation.
,,-L O
I 9
TP 800.04.1 7.2.7 Itintain EG IcVel at 100" on the SUR using the Aux n1 p g s, VerJ.fy tik t uotn cx3 S/U anc Main Feecanter Valves are shut.
S-it, 3
lA'3 the SUFP then bring on the Aux Boiler and shift the Isux steam he.
g7,.,
to the Atr: boiler.
Shift the Ihin R1 pump to the Aux steam heas' Rp; and shift gland steam to the Aux Steam header. Decrease Main E1 A.!
pump as dJrected by the Shift Foreman.
,?. -
Verified Date 7.2.8 Allca ter trature and pressures to stabilize and reverify the follcaing i.nitial conditions:
Peactor Poser 2-4% FP (Correcto3*)
Xe Ccndition (As specified in step 6.2.4)
Doron concentration (As specified in step 6.2.15)
Pressurizer L2 Vel 100 - 150" 2 ID Ccolers in service Corr.ct indicated paer for temp 2rature using correctica dete.:..ined in Phase I (See Attac' rent 8).
t Verified Date 7.2.9 Defeat Ic :s of fica trips in tha Paactor Protective Systc;n by cagletin, Attachment 5.
(
Verified Date N
7.2.10 Using dig cal voltmater measure carpensated locp ficw voltage in INI at cac:.ensated ficw nultiplier cutput.
N'1I IDCATICN PIN
~
M.
ILCP 1 7-6-5 7
IDOP 2 4-3-9 7
i' Record 5 nd.nutes of fica voltages at 30 second intervals. Eccorc c~
c, voltages o.. Attachment 9.
i Verified Date 7.2.11 Begin trending canputer data as spe:ified in Attachmont 4.
Verified Date l
7.2.12 Trip the running reactor coolant p:nps. after observing the tbtes of this step and step 7.2.13.
IDrE: Flow coastdcwn time will be approximately 30 seconds.
As fica coasts dcwn, reactor o T will increase and should stabilize around 30-40 F. with flow sustained by natural circulation. With negative mcderator coefficient, reactor poeer should be stable, but respcase to temperature transients will be sluggish due to loa ficw.
i Prim ry temperature will tend to increase due to the reducal ficw,,with, corresponding increase in pressuricer
.1 c :n1.
p g
IV)Fm
- o TV aAJ m
2310 251 fJa
i 10 TP 800.04.1
[
7.2.13 Carefully nonitor RCS tcr.peratures, pressures, pressurizcr IcVel E/..
and reactor poaer. Minimize the effects of the primary tctgeratur f.
increase by over feeding the steam generators and allowing level to increase to a maxinun of 50% OR 'Jhe primat'y pressure transient may also ha mitigated by increasing letdown flow.
s 1
Manually trip the reactoi and start two RC prnpsin each icop if auxiliary feedwater ficw is lost or if any of the following limitt are reached:
PARAVETER HIGII LIMIT 104 LD11T Irdicated Reactor Power (NI5,6,7 or 8) 5% FP
l l
Pressurizer IcVel (LPS PCl4) 280 IN.
50 IN.
OISG Pressure (PI SPRB cr PI SP12A) 1020 PSIG Any Incore T/c ('Isll - T562) 650 F Th (TI RC3B1, PC3A1) 600 F
(
CST Icvel (Tank in Service) 10 Pr.
[-
RCS Tave 525 F Corrected (See Section 7.1)
Verified Date t
j 7.2,14 Move control rods only as required to keep. reactor p:r+er between 1 & 4% FP (Corrected).
Verified Date 7.2.15 Allow Th, Tc and PCS pressure to stabilize. Verify that* natural circulatic;n has begun by observing that reactor 6T has stabilizec in each loop. 'Ihis may take up to h hour. Expected natural circulatica ficw is on the order of 7% which will result in a reactor 6T of about 30 F.
\\
Verified Date 7.2.16 Gently adjust reactor pcwer and steam generator level to achieve stable conditions at 40% OR level and 2-4% FP corrected.
NOIE; Intenysliate Range level may provide a nure accurate scale 1
for controlling reactor pcreer.
[
Verified Cate
(
7.2.17 Record data for at least 10 minutes after reactor AT and reacter inaer have been stabilized.
Verified nate 23'0 252.._
m 6
11 TP 800.04.1 I~
7.2.18 Shif t Ian ] Turbine Dy;oss Valve to !!ND (PIC ICS 12D) and o;cn it
- 57. to reduce
'Do not excccd 100 F/llr. ccoldom,1-1 OrSG pressure by 50 PSIG.
rate (
20 psig/ min.)
$'i NOTE:
If the Pc.ctor is trippd while a 'Ibrbine Dypass Valve is in
-h{
HAND, ret'en the valve to the CIOGED position and return to Af1IO t
, if directal by the Shift Fore:ran.
t Verified Date 7.2.19 Mani'.or B: ush Recorder traces. The tcs:perature drop should be
+.
h detected ca the Icop 1 'Ih trace in about 1 minute.
a.
I' Verified Date I-7.2.20 Continue to maintain reactor pcwer 2-4% FP corrected using
.~
control rods in Manual.
i Verified Date
~
7.2.21 When the r mperature drop is detected en the Icop 1 di Trace, shift Icc : 2 Turbine Bypass Valve to IWO (PIC ICS 12A) and open 1
it/v5% ta reduce 1-2 OTSG pressure by~50 PSIG. Do not exceed 100 F/llr oldoen rate (N20 psig/ min.)
IFXE: Do at alloa Tav to go telcw 525 F.
Verified Date 7.2.22 Continue --king data until the tc=perature change is detected on the Io:o 2 Tn Trace in the Brush Recorder. This should +d e approxiral ly one minute.
Verified Date 7.2.23 Pcturn Turbine Bypass Valve H/A statiws in AUIO ad alloa Tav to increase back to equilibrium value for turbine header pressure of 870 psig. This will be about 534 F.
Verified
_Date 7.2.24 Clear the t; rip on the SFRCS Panual Trip ITIS 4369E & IIIS 4870E by releasing the hold-in trip buttons. This will alloa the SFPCS to reset when RC pumps are started.
Verified
_Date 7.2.25 Panually trip the reactcr. Verified Date
.l D
~, D
- D ~@fM
. _ _ - - _ _ _ _... 2 3.10. 2.53.c
L
-f 12 TP 800.04.1 t L.5
,'-.C 7.2.26 Start te RC puqsin eacn loop, then verify Auxiliary y
$1 Feed Purp Mcde Switches IIIS 5203 & HIS 521D (C5709) in Ws?.ML arx1 back Auxiliary Feed Puups dan to muumn spoed.
3,,, e -
O.
Verified Date u
M' p
7.2.27 Carry out Reactor Trip Procedure, EP 1202.04.
~
Verificd Date 7.2.28 As steam generators steam dm.n, verify S/U Feedwater Valves cpan 1
and control on Ioa Invel Setmint.
Bring the running P.ain m p:
up to specd and stop the SUFP as directed by the Shift Fore: an.
Verified Date 1
7.2.29 Paturn th2 Auxiliary Foodwater System to Standby Creration per SP 1106.06, ;suxiliary Fecdwater System.
i Verified Date 7.2.30 Stop taking data on reactimater, recorder and carpater.
Verificd Date e
7.2.31 Rcmove jugars installed cn RPS in Step 7.2.8.
_l Verificd Date (TS) 7.2.32 Recet high flux trip as directcd by PES Test Coordinator.
('IS 3.10.3)
Verified Data 7.2.33 Infonn Shift Forcrnan that this portion of the test is ccnplete,
=
n.
7, Verified Date 7.2.34 Calculate natural circulation ficw per Attachment _6.
Verified Date 8.
ACCEPTMCS CRITERIA 8.1 Natural circulation ficwrate determined by eith2r mathed in Attach ent 6 eqtnis or exceeds the minin:m specified by Attachment 7. '
Verified Date
/..
9.
RESULTS DISTRIBUTICN 9.1 B & W Site Operations Panager 9.2 TEro Pcwer Engineering & Construction Division
(
'?
e S
en
1..'
13
- f f
REACTI!!ETER DAT's FOR MIASE I
'I? 800.04.0 4[
m
~
Ict up the reactimeter to record the following data:
SIGNAL
{
PARAMF.TER RANCE' RANGE __
.SITTP MO.
g
- w. e i Conarated IG 0-1000 int 1-100 tiv 22 6";
S.
I, 11U Tank Level 0-100"
-10 + 10 VDC 35 4.
Int. Range Flux (NI3) 10-ll-10~3 0 't 10 VDC 40 h.. l, '
a
. 10-3 0- +-10 VDC 41 f, ',
Int. Range Flux (NI4) 10 a
Power Range Flux (NIS) 0 - 125% FP 0 - + 10 VDC 42 l..:
Power Range Flux (NI6) 0 - 125% FP O - + 10 VDC 43 Power Range Flux (NI7) 0 - 125% FP O - + 10 VDC 44 Power Range Flux (NI8) 0 - 125% FP O - + 10 VDC 45
_ Loup 2 Th 520 - 620 F + 10 VDC 50
!~
I Loop 1 Th 520 - 620 F + 10 VDC 51
,- 3 Press. Level (Comp.)
0 - 320" + 10 VDC 52
- i 6
NR RCS Pressure 1700 - 2500 PSIG 0 - + 10 VDC 53 -
1 Loop 2 Tc (1-2-1) 520 - 620 F + 10 VDC 58 1
Loop 2 Tc (1-2-2) 520,- 620 F + 10 VDC 60 520 - 620 F + 10 VDC 62 Loop 1 Tc (1-1-l)
Loop 1 Tc (1-1-2) 520 - 620 F + 10 VDC 64 6 #/Hr
-10 ' + 10 VDC 70 S/U W Flow Loop 2 0 - 1.5.10 6
3~
S/U P.i Flow Loop 1 O - 1.5.10 #/Hr + 10 VDC 71 EN W Flow Loop 2 0 -- 7.106 #/Hr + 10 VDC 75 -
I FCI W Flow Loop 1 0 - 7.106 #/Hr + 10 VDC 76 Feedwater Temp.
0 - 600 F + 10 VDC 74
' Croup 6/7 Positien 0 - 100 7.
0 - 5 VDC 3
i OTSG 2 Outlet Presa.
0 - 1200 PSIG t 10 VDC 72 911n or Ki ;;,o t' -
LJlU 4J
.O g
e
... t 14 TP 800.04.1 THE TOLEDO EDISON COMPANY;
'~-
POWER
- PRODUCTION DEPT.
3
.totior.
TEST READINGS Sheet No.
1 of 2 __
kr5 I OF cst of Par,e No.
4 :. 1. rest No.
Observers l 'hte
~
9.
l rice T V[
f t-,- Yf 3-g[" ' "K~
- ='I
- Q ~ - - ~ ~==
1 Int
,60,03 HEAT BAIATE l
I l IDOP 1 IDOP 2 IDOP 1 IDOP 2 GEN PIANT p/U B4 S/13 B1 FN Bi FN FW Mi l
CQGUTER I
IFIU1{FIci FIDi
- FIDW i
FUNCT.9
'R!
I., Timeq __ __._ -
_m p_ _._,
. _ ___J _
I GOU.P 2. 2..__
i.
1.
. 7 ____..
j j l
l
~i~
r t i i
1
-i
!. i-i I
i i
=
i i
i I
l 2
- 4. __ _. __
i t I
{___
l 2 _.
Ii-1 i
I I
l
.l T~~
I
~
l i
j l
l l
l l
I i
j.
I
_1_
l i
i l
I i
I I
i i
l l
l I
l'
~*
I T
s I
j l
e 1
l 8
I t
i t-1 T_
i i
l l
1 i
l
-f i
i i
i I
i
]'
{
i 1
I I
I i
.I i
8 I
f
- __L, i
i i
i i
e i
l 1
i j
t.
,I.._
1 i
, l h
b I
f I\\
2
~
'1 1
i t
a g
__y j...
j.. _._p _._.;......j_,
2310.2~56
,y C. -
ID oC27 15 TP B00.04.
THE TOLEDO EDISON COMPANY' c.
f-P.
POWER
- PRODUCTION DEPT.
in TEST READINGS Sheet No. 2 of 2 f"', Test of--
Pa ge No.
<r 3..
rest No.
7',
If Tave...d.ro.p.s.. b. e..l_o__w 5_.32 F o
Observers Nor3:
g.,
g
[#'
rine
_r_Scord_all_ reading.s at 15 Min. intervals.
.c._,--
_,m_-===.=_
_ __,m _ m -_=,_ __
_.mr.__.--
_-.___.e==
r,-.
1 I
i Int. No*NI-3 p..
p NI-4 'NI-5 NI-6 NI-7 NI-8 Tr4Bl TI4B3 Tr4A1 Tr4A3 I
Int.
jRange.Pange. Range Int. ~ Pwr. : PW.
Pwr.,Pwr.
ILOP 1 ILOP 2 Pange Bange Range NR Tc NR,Tc g
1 Tirne Flux i Flux i l
t o!
O
- -,(A_mp_s. )_..( A.m. ps_)m1%, FP_),_(%._,m ) m(% F.P_._.(.% FP) p!
gg i
p FF i
.__ )..__ -. _.-_ __ _ J __ _ _ _x=m.-- = =!
.m, l
==_m-4._
._.L.____..
i i l
I i_
i
__ e.___
l i
I 1:
.l l
l l
1 t
i i
I j
i i
i I
' L i
J l
l t
g e
q I
- l..
i 1
3 i.
j_. _..
I j
.i i
I I
l 1.__
I I
i J
i 8
t i
i e
i t
t t
i i
e t
i i
I i '
I l
l l
r i
i j
_i.
l 8
I e
I i
l l
l l
l I
h 1
l i
1 I
i e'
i
-]
I t.
b i
i 1
l
}
i i
i l
1 I
-T a
1 i
i j
)
I ~~
g-i l
i s
g' l
1 1
l 3_______
._f.._
N
'f
~~
l g
1..
_...h_____.
,...._.t._....._l__.____.____......
..g --
j q
- _ 4........ - L __... _. j.._. l
-.4...
N-
~
2310 257 N
C 1
f.J 16 TP 800.04.0 REACTDtE. ER & RTORDER DATA FOR FHASE II T
.7 If" '
Set up the reactheter to record the following data:
i&.
g.
PARAMETER SIGNAL RA?E
_ RANGE SUrP NO.
PCS Carpensated Flow Icop 1 0 - 90.106
,. + 10 V 48 BCS Cor.pensated Flw Icop 2 0 - 90.106
_lo _ + 10 y 47 PU Tank IcVel 0 - 100"
- + 10 VDC 35 power Range Flux (NI 5) 0 - 125% FP 0 - + 10 VEC 42 Power Range Flux (NI 6) 0 - 125% FP 0 - + 10 VDC 43 Pcwer Range Flux (NI 7) 0 - 125% FP 0 - + 10 VDC 44 Pcuer Range Flux (NI 8) 0 - 125% FP 0 - + 10 VDC 45 Icop 2 'Ih i
520 - 620 F + 10 VDC 50
.Icop 1 Th 520 - 620 F + 10 VDC 51 Press IcVel (Carp.)
0 - 320"
...r + 10 VDC 52 NR PCS Pressure i
1700 - 2500 PSIG 0 - + 10 VDC 53
~
Icop 2 Tc (1-2-1) 50 - 650 F + 10 VDC 59 l
Icop 2 Tc (1-2-2) 50 - 650 F - + 10 VDC 61 Icop 1 Te (1-1-1).
SO - 650 F
' 10 - + 10 VIE 63 Icop 1 Tc (1-1-2) 50 - 650 F + 10 VDC 65 Croup 6/7 Position 0 - 100%
0 - 5 VDC 3
OrSG 2 Outlat Presc 0 - 1200 PSIG
\\'- + 10 VDC 72
\\;
OrSG 1 Outlet Press 0 - 1200 PSIG + 10 VDC 73 OISG 2 Operate IcVel 0 - 100%
i
. + 10 VDC 77 I
OISG 1 Operate IcVel 0 - 100%
' + 10 VDC 79 JISG 2 S/U IcVel 0 - 250" i + 10 VDC 78 OrSG 1 S/lJ Invel 0 - 250"
~10 - + 10 VDC 80 J'
2310 258
..e
,,.... =..
O e
e
E 1
17 TP 800.04.0 k.
REACTIFETER & REORDER DA2A FCR PHASE II h,.-
i 2.
_ Set up Brush Recorders as folloas:
A.
Recorder No. 1 t
TRACE PARAVETER RI4E SUTP I 1
Icop 2 Th 520 - 620 F 50 2
Icop 1 Th 520 - 620 F 51 t
3 Icep 2 Tc (1-2-1) (WR) 450 - 650 F 59 1
4 Icop 2 Tc (1-2-2) (WR) 450 - 650 F 61
~
5 Icop 1 Tc (1-1-1) (WR) 450 - 650 F 63 ti Icop 1 Tc (1-1-2) (WR) 450 - 650 F 65 i
NOIE:. This recorder must be positicned in the Control Pccra so ttat the Test leader. can see -it.
2310 259 N,
=
1
(\\
.s ATTIGE7F P/GC 2 CF
- N
~~
18
'1? 800 04 0
- Cn @trra1 DATA - PHASE II tt 1.
Program an alarm point for incore thennoccuple tenperatures T511 - T562 in th plant cmputer with setpint 645 F.
~ ~..
2.
Place the follcuing para:reters cn. trend recorders on the Operator's Console:
- v -
il,.
Icop 1 Th T 721 Icop 2 'Ih T 730 f.'.
Icop 1 Tc (hn)
T 781 Icop 1 Tc (Wa)
T 801 Icop 2 'It (va)
T 821 Icop 2 Tc (WR)
T 841 Icop 1 Cmp. Flcw F 727 '
Icop 2 Ccnp. Flow F 732 1
l 3.
Place the follcwing incore thennoccuples on the line printer to print at l
1 minuto intervals:
CORE IOCATICN Pr. ID. NO._
H-9 T 541 F-7 T 531 M-9 T 542 L - 11 T 552 E - 11 T 549 L-3 T 515 j
P-6 T 528 l
C - 13 T 556 0 - 12 T 555 1,
B-7 T 529 E-4 T 517 H.- 5 T 521 L-6 T 526 G - 11 T 550 E-9 T 539 G-2 T 512 N-4 T 518 R-7 T 533 i
N-9 T 543 C - 10 T 544 4
t-
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(refer to the next age for core trap.)
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20 TP 800.04.l r, ;.
~
PRCCEDURE FOR DEFEATHK) RPS TASS CF FIC1 TRIPS FOR NA'IURAL CIPCUIATION TEST p; ~ ~ -
- p...
.j 1.
O tain keys for RPS Cabinets frcm the Shift Forcrran.
2.
Defeat Powr/ umps, trip by installing jumpers as follcws:
?
JUMPER JU'GER RPS OFREL ILCATICN PEG R;STNJFD/DATE RDC/ED/DATE 1
1-3-8 17 - 18
/
/
2 1-3-3 17 - 18
/
/
3 1-4-P 17 - 18
/
/
4 1-4-ft 17 - 18
/
/
, 3.
Defeat Flux /Flcw trip !.y installing jt n,mrs as follcus:
e-
~)
JUMPER JUMPER RPS CHANNEL I/X'AT:]
PD;S E1STAILED/DATE TO U/ED/DATE 1
1-4-14 17 - 18
/
_/
2 1-4-14 17 - 18
/
/
3 l-5-14 17 - 18
/
/
4 1-5-14 17 - 18 ~
/
/
4.
Icck BPS Cabinets and return keys to Shift Forcran.
~
'l Verified Date 1
I
~
2310 262 t
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21 TP 800.04.0 CALCULATION OF NATURAL CIRCULATION FLOWRATE iif
- RIf..
?
METHOD NO. 1 - REACTOR $
- m.,.
This method is based on the relationship between primary flow, heat input & entha:
q (BTU /HR) = Wa 6 h, and ast umes that for the range tested, o h = h T.
i 1.
Average compensated 1 rep flow voltages (Vm) which were recorded on Attachment Using the average voltcge for each loop, co=pute compensated loop flow (Wf):
~
Vm + 10 Vm + 10 I
^
6 Vm 20~~
Wf = (
20 ) (90.10 ) lbejhr @ 6080 Th L00P.1 LOOP 2 2.
Delog reactimeter and.!ve. age 1 min. of RCS temperature data just prior to RC pump trip (Step 7.2.12).
3.
Using these averaged t.a peratures, determine reactor A T jusc prior to RC pump trip ( 6 Tf):
A Tg = Th (l'r...+ Th (2)
Tc (la) + Tc (1b) + Tc (2a) + Tc (2b) 2 4
o F
=
4.
Using the reactimeter data taken after steady state natural circulation has been established (Step 7.2.17), average 10 minutes of RCS temperature data.
5.
Determine natural circul.ation reactor d T using the data averaged in Step 4:
l ATn. Th (1) + T h (2)
T (la) + Te (lb) + Te (2a) + Tc (2b) c 2
4
=
F.
6.
Using reactimeter data for time just prior to RC pump trip, average the power range levels for the 1 nin. prior to the pump trip.
NIS NI6 N17 NI8
% FP 7.
Determine the average power level prior to pump trip:
(
\\s p
NIS + NI6 + NI7 + NI8 4
g 8.
Using the same period of time as step 4 above, average the power range levels for the 10 minute period:
8 3.s NY(
?!T C MT7 S'"
r
.m
C/.LCULATION OF NATURAL CIRCULATION FLOURATE 9.
Using the correction factors developed in Section 7.1, correct the power rang
. '],-y levels in Step 8 to the Te Prior to the pump trip.
NIS NI6 NI7 NI8
% FP' i
~
- 10. Using the corrected power levels in Step 9, determine the average power level after natural circulation flow is established:
% yp NIS + NI6 + NI7 + NI8
=
p e
4 11.
Calculate natural circulation flowrate:
Vn (lbm/hr) = Wg (Loop 1) + Wg (Loop 2),
12.
Calculate natural circulation flowrate as % of 100% flow at 6080 :
F Wn (100) 71oy 131.100 METHOD 2 - INDUCED TDiPER/.TURE TRANSIENT TIME
\\
This cethod was a direct measurement of the time for a te=perature perturbation to travel from the T instrument to the Th instrument.
c 1.
Plot the following reacti=eter data taken during steps 7.2. 18
-7.2.22 versus time:
. LOOP 1: Th (1), Te (la), Tc (lb)
LOOP 2: Th (2), Te (2a), Tc (2b)
~
2.
From the plots in step 1, determine the time between the temperature, drop at the T instruments and the Th instruments in each loop.
c (NOTE: This time should be on the order of 1 minute.)
6t (loop 1) =
Min.
~
At (loop 2) =
Min.
3.
Calculate the natural circulation flowrate using each loop 4 c: -
44,040 (Cal / min)
At (mtn.)
Un" Ae LOOP 1
(
LOOP 2 Where 44,040 - Volume of RCS (in gal.) between Te&Th instruments.
I 2310~264
? +:. j '.
23 TP 800.04.0
- f.L.
CAI.CULATION OF NARJRAL CIRCULATION FLO*. IRATE
=$ ie '.
y 4.
Average the flowrates obtained using Loop 1 & Loop 2 8 c:
j d
g.,
Wn (1) + Vn (2)
.{.
Wa (av.)
=
GPM
=
- $1 > 1 2
p.
?:.
T.
5.
Determine natural circualtion flow as a % of rated flow:
n.
(.y. '
~f f..
W (av.) (100).
g fio, n
352,000 gpm I
i l
we 2310 265 o
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La 25 TP 800.04.0 r,
~, i -
DEIERem1ATIQ1 OF CORRIX?rICt1 FACIOR IVR INDIC/CED PO.ER 4
_,, 1.
Using reactincter data frcm Section 7.1, select tonperature & NI Power Data for each tanperature plateau of Section 7.1.
De sure to select data where This has been power level by heat ba..ance and feedwater floas are constant.
3.f C l
" marked" on the reactir. uter data file by starting a new block of data when fg it.. !
- conditions have stabilized (See step 7.1.11).
s, V.
For the selected data records, average the four 'It values to obtain an average E
2.
7 Tc at a given tinu.
3.
Plot the average Tc for each plateau versus the corresgx4 NI3, 4, 5, 6, 7, 8 flux levels:
8 Indicated Power VS.
Avg. Tc for NI-5, 6 NI-nPD e
j Indicated GP 6/7 Pcw_r NI-6 i
Boron Conc.
t
]
559 582 BC Press Tc F
g
\\
1
~
4.
For each detector in Step 3, linearize tha data and find the slope of the lin:
witich will be the conection faci.or forf that dett:ctor.
5.
Using the correction factors frcm step 4 atcve, construct charts for operators use shcwing correction to be applied to indicated paur V. 'Ib:
a.
Fake chart similar to this. Since
-0.2
,. -0.1 all detectors should be approximatc
/j ly the sa:ne, it should be possible C.F.
G 40 1 y'
to make.one chart for all pcwer
+0.2 range detectors and one for inter-nediate range.
l
-2
-1 0
+1
+2 bTc 6.
The correction factor will be used in Phase II to correct irdicated per:
t 1
P
- Ind. + CF Acttul
=
I k
2310 267 j
1 I
e
r-r
?G l
e- - n TP 800.01.0
'e i
'f.
THE TOLEDO EDISON COMPANY c.
I 'd*
POWER PRODUCTION DEPT.
,," Station.
TEST READINGS Sheet No.
-1 Of 1 OF Test of.
Page No.*
e..j Test No.
Observers x.
Date
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Time.
f.'
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- D SOM #055 l
July 16, 1975 L *-
s Page 2
(..
D}b6 fb W. J. G. h ans 73 dabcock & Wilcox Test Results Gu2=sary:
Natural Circulation in Oconee I T70e Plants, "A Description of Measure =ents and Su:::ary of Results' e
[
March 21,1974, as evaluated by BfB Nuclear Service. Natural circulation steam generator level approxicately 50% on the operating range instru=entation.
Attach =ent 2: Natural Circulation L
^.
Graph 1:
Flow vs Reactor Pcver (Decay Heat) @ Natural Circulatien Level of 50% on the operating range instru=entatica " Conservative analytical results".
l Graph 2:
Same as Graph 1, except curves represent " Realistic analytica; results".
These two graphs represent our Engineering Department's evaluation of charact istics for Davis-Besse Unit I as cocpared to the Oconce I natural circulatic:
characteristics normalized to Davis-Besse pcuer and ficv pa aceters.
Graph I represents the conservative characteristics in which all heat transferred frc the Reactor Coolant System to the Secondary System is assu=ed to occur in the g.
lover portion of the steam generator, in the vicinity of the lever tube sheet i
Graph 2 reprezents the more realistic situatice for it assu=es that all heat transferred frem the Reactor Ccolant System to the Secondary System occurs ir
{
the vicinity of the auxiliary feedvater nozzles where feedvater is introduce:
to the steam generators. Both graphs are based on a natural circulation leve I
of 50% on the operating range and as indicated the more realistic characteric f
tics, Graph 2, reflect more natural' circulation flow than the conservative rc r
sults shown in Graph 1.
.ach case the natural circulation flov at Davis-Besse Unit I vill always b e,reater than that at Oconee I, a plant of like design, for the corresponding power conditions.
I hope that this information is sufficient for deletien o f the Natural Circu tion Test and for making the necessary corrections to the FSAR.
If you have questions, please contact Fred Faist.
h.
\\
Yours truly, ha-R. J. Baker, Jr.
Site Operations Manager fi RJB:FRF:nif encl.
cc:
G. M. Olds E. C. Novak, TECo E. J. Coppola J. D..Lenardson, TECo J. A. Lauer 1.
T. D. l'urray, TECo R. t. ritt=a" 2310.269 0 T. F. Scott
~
E. L. Logan E. R. Michand...
4
~7 I
L.
7/it/75
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tt-Natural Circulation in Oconee i Type Flants
-t l,
A Description of Measurerents and Sum.ary of Results
.Harch 21, 1974 d
~ ~. _.
- }7] O 2 7.,
3
~
I
,~
l -
a g.-
m 4.
f I,
[
,(
Prepared SY:
s i (
A. Robeson o # e 4 i e Plant Per formance Service l
i l
Reviewed for Acc.uracy:
7
,_._.g____
e
~
NATURAL CIRCULATION IN OCONEE I TYPE PLANTS i
s Sum =ary_
dMl Adequate natura1' circulation in the reactor coolant system assumes heat.recoval from the reactor core upon loss of all reactor coolant pumps.
Babcock & Wilcox I
nuclear steam systems are designed to provide natural circulation, and safety I
analysis verifies that in the natural circulation mode, more than adeqt; ate
~
l cooling is provided for the reactor core.
As indicated in the Oconee I FSAR, the systee is designed to provide natural circulation flow, at 1% decay heat, greater thar. that required for heat removal by a factor of five.
Natural circulation tests performed on Oconee I, using two independent methods, yielded a minimum factor of ten.
Although normal reactor coolant flow sensors are not intended to read flow rates in the range produced by n'atural circulation, a flow rate in agreement with test measurements was indicated in the control room.
Thus, analysis and experimental measurements have verified that Oconee I-type plants are capable of adequate natural circulation flow in the reactor coolant system upon loss of all reactor coolant pumps.
t Method of Measure =ent 4
i ne bar ts for measurement of natural circulation is determination of the reactor core transit time for a temperature transient, the " induced temperature transient
' circulation time (ITTCT).
Decay heat from the reactor core provides the flow, and', prior to the measurement, the reactor is operated at power for a time which vill insure that at least a 1% full power decay heat level will be present during the first hour following reactor shutdown.
The.once-through steam generators (OTSG) are operating at a level of approximately 5_07., and the reactor is brought to hot shutdown condition with one reactor coolant putp (RPP) operating in each loop.
The main feedwater pumps are stopped and the level in the OTSC's mair.tained ar. 507..by the emergency feedwater pump through the auxiliary nozzles.
The remaining two RCP's are tripped, and the core outlet temperature allowed to level off, indicating a stabiliz'ed natural circulation flow.
6 j The time required is about one-half hour.
When natural circulati.on is established, a temperature
-t drop of about 10 degrees is produced in.the core inlet temperature from a rapid reduction in header pressure by opening the steam bypass valves.
A measurement
(~
of-the time between the break point in temperature at the cold leg.and that at the hot leg, determines the circulation time.
t The volume and weight'of water between the two temperature measuring points and the observed transit time are the parameters necessary to determine the. natural circulation flow (NCF):
h g.
u cp 2310 271 (1) uCr
=
(
where:
M Mass of the reactor coolant between cold'and hot leg
=
RTD's at the time the ITTCT is measured in Ibm.
ITTCT Induced Temperature Transient Circulation Time in hours.
=
~
~
e
i V:6' i
=2~
W.
F
. An alternate value for the natural circulation flow is obtained by using a calculated value of the decay heat source in the reactor at the cice of introduction of the temperature breakpoint.
Using a calculated decay heat
[
curve and the recent power history of the plant, a value for the decay heat yJJ.
generated by the core for any time af ter shutdown can be obtained.
g t
The calculated value of Q, the decay heat generated by the core (Stu/hr), permit Q( -
the natural circulation flow (NCF) to be calculated by:
1 I
-)
I 0
NCF
=
C x (Th-T)
(2) where: Q CalcuJated value of decay heat at tice of ITTCT ceasurement.
=
C Specific heat of reactor coolant in Btu /lb F
=
Th Temperature of hot leg at time breakpoint occurs at hot leg.
=
T Temperature of cold leg at time breakpoint occurs at cold leg.
=
c
. Experimental Data
- The initial experiment
' conducted at Oconee I on November 4,1973.to measure natural circulation flow with dec The results showed that there was adequate natural circulation, but due to problems with data retrieval, no accura:
,)
l value for ITTCT could be established.
A second experiment, on May 2, 1974, yielc 7
the results shown below.
~
Natural circulation measurecent using Equation (1):
M Mass of reactor coolant between cold and hot Icg RTD's
=
~]
254,541 lbm
=
1 min. (Graph 1) 2310 272 IITCT
=
NCF=(254'541 6
x 60 min /hr = 15.3x10 lbm/hr @ 1.05% Decay Heat
[
The pressure transient which produced the te=perature drop is shavn in Graph 2.
The sharp temperature decrease (breakpoint), used as a timing indicator to
.aensure ITTCT, produces an increase in natural circulation flow which is included in the above value of NCF.
A rough correction can be made by using the results r
of a similar experiment performed at zero power, beginning of life (May 1, 1973) when the decay heat was essentially zero.
Graph 3 shows the result of a ceasurect i
yielding an ITTCT = 4.0 min.
The flow rate corresponding is rhen:
NCF (No Decay Heat) =
256'947 6
x'60 = 3.85x10 lbm/hr 4O
~
- Pressure and temperature conditinna vara nae
'8-**---'--
t
(-
2
~ ~ ~ ' ' ~ - ~
~
1
e l
'[
^y
?
3-b g
From the FSAR; Oconee I (1% Decar Heat Power):
6
~
2.34 x 10 lbm/hr Calculated NCF Available
=
0.47 x 10 lbm/hr Iculated NCF Required for Heat Removal
=
Using Equation (2), an alternate value for the natural circulation flow can be determined:
i 9
7 q = (0.0105)(8.767x10 ) = 9.21x10 Btu /hr NbTE:
8.767x10' Btu /hr = 2568 }Gt C = 1.22 Btu /lb F T
-T 25.5 F
=
h 7
9.21x10 6
NCF = 1.22x25 5 = 3.4 x 10 lb=/hr
-l Conclusion'
~
}
Experi= ental mecsurement of natural circulation in Oconee I type plants has j
shown flows more than adequate to ensure heat removal from the reactor core upon loss of all reactor coolant pu=ps.
4 Although both of the methods used to determine natural circulation flow yield more than adequate rates of flow, the value obtained from Equation (1) is considerably greater than that obtained from Equation (2).
The writer feels that both values are representative, but has a somewhat greater confidence in the measured value [ Equation (1)], for the following reasons.
j j
i
'Ihe experiment is basically simple, but fluctuation of parameters on the secondary side may make it dif ficult to obtain flow equilibrium or determine precise temperatu averages.
As shown in Graph 1, even with the scatter in data points, the tecperatur I
breakpoints can be located with reasonable precision.
The secondary side was fairly y
stable during the measurement.
Since the mass of coolant involved can be deter =inec with accuracy, the value of NCF prior to correction should be quite good.
The i
correction applied was based on results of somewhat lower reliability (Graph 3),
the correction is clearly demonstrated.
l h.e method employing Equation (2) depends on a measurement of the temperature drop across the core, which in this experiment is known reasonably well, but is certainly not a constant value.
The value of Q requires a!: detailed knowledge of the power
}
history, and dependence on properly interpreting the power history in terms of a calculated fission product decay heat curve.
In the 1% decay heat region.,the
.. decay _ heat drops by.a factor of five in about one hour. so an accurare kn,culedce
~~
,.._a
.