ML20091K822
| ML20091K822 | |
| Person / Time | |
|---|---|
| Site: | 05000000, Cooper |
| Issue date: | 08/31/1982 |
| From: | Laudenbach D EG&G, INC. |
| To: | NRC |
| Shared Package | |
| ML17139C189 | List:
|
| References | |
| FOIA-84-105 NUDOCS 8406070213 | |
| Download: ML20091K822 (7) | |
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s August.1982 DRAFT TECHNICAL. EVALUATION OF THE ELECTRICAL, INSTRUHf Nl ATION, AND CONTROL DESIGN ASPECTS OF THE PROPOSED LICENSE AMENDMENT FOR SINGLE-LOOP OPERATION Of COOPER NUCLEAR STATION d
(Docket No. 50-298) by D. H. Laudenbach EGAG, Inc.
Energy Measurements Group San Ramon Operations 8[406{0213840319 PDR Lg4.- 05
s; s
I.
INTRODUCTIUN By letter to the U.S. Nuclear Regulatory Cosmiission (NRC) dat.ed August 5,
1980 [Ref. 1), the. Nebraska Public Power ~ - Di st rict submitted informat. ion to support. its propo' sed license amendment 'to7perate the Cooper Nuclear St.ation (CNS) with one recirculation loop out of service (i.e.,
single-loop operat.f on).
1his inicrmation included the licensee's analysis of significant events, which were based on a review of accidents and achnormal operational transient.s associat.ed with power operations in the single-loop mode provided by General Electric
- Company, Nuclear Energy Division (GE-NED) the nuclear steam supply system designer.
Conservative assumptions were employed, as discussed in GE-NED report NEDO-2425S dated May 1980 [Ref. 2], to ensure that the generic analyses for boiling water reactors (BWR 3 and/or 4) were applicabic to the Cooper Nuclear Station.
In response to a request for additional information, the licensee provided supplemental information in a letter dated May 6, 1982 [Ref. 3].
Subsequently, two telephone-conference calls were conducted with the licensee [Refs. 4 and 5] concerning protection system trip point setting changes for CNS single-loop operation, and documented by the licensee's lett.er dated July 28,198P [Ref. 6].
The purpose of this report is to evaluate the electrical, instru-mentation, and cont.rol (EIAC) design aspects of the pro)osed license amend-ment change to the CNS technical specifications.
T ie consideration of proper plant variables, computer models, and the licensee's conclusions on core perf ormance and clad temperat.ure are outside the scope of this evalu-ation.
lhts review was conducted using IEEE Std-279-1971 [2ef. 7]; NRC Branch Technical Position ElCSR-12 [Ref. 8); the Code of Federal Regula-tions Title 10, part 50, Appendix A. " General Design criteria for Nuclear Power Plants" [Ref. 9]; and NRC Review Criteria detailed in Project 8 of FIN /189A No. A-0250 [Ref. 10].
II.
EVA1.UAT10N AND RECOMMENDATIONS The licensee indicated that current CNS technical specifications do not allow plant operation beyond a relatively short period of time if an idle recirculation loop cannot be returned to service, in the event maintenance of a recirculation pump or other component renders one loop inoperable, the capability of operating at reduced power with a single-recirculation-loop is highly desirable from a plant availability / outage planning standpoint.
The licensee's proposed technical specifications would allow the racctor to operai.e in single-recircul ation-loop operation for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> bofere making any setpoint changes to the reactor protection system.
With I
cne recirculation loop out of service for greater than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, the reac-tcr would not be operat.ed at a rated thermal power greater than 50%.
In crdIr' to continue to operate the reactor in single-recircul ation-loop c uration beynnd 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, it will be necessary to make setpoint changes to' t o SCRAM trip sett.ings of the average power range monitor (APRM) system I
1-
and to the rod-block set. tings of the rod' block monitor (RBM) system.
Because of the dif f erent flow rate arid path during si ngl e-recirculation-loop operation, the APRM SCRAM trip sett.ings, which are flow-biased accord-ing 1.o the equation in the proposed technical specifications, require reset.t.ing to protect the reactor from overpower.
The rod-block setpoint equation is flow-biased in the same way and with the same flow signal as the APRM setpoint., and must al s~o be modified to provide adequate core protect. ion for a postulated rod withdrawal error.
i The licensee provided the following technical specification bases for the APRM SCRAM trip settings:
The average power range monitoring ( APRM) system, which is calibrat.ed using heat balance data taken during steady state conditions, reads in percent of rated thermal power (2381 MWt.).
Because fission chambers provide the basic input signals, the APRM system res-ponds directly to average neut.ron flux.
During trans-I ients, the instant.aneous rate of heat. transfer from the fuel (reactor thermal power) is less than the instant-aneous neutron flux due to the time constant of the
,i fuel.
Therefore, during abnormal operational trans-l ients, the thermal power of the fuel will be less than l
that indicated by the neutron flux at the scram set-ting.
Analyses demonst rate that with a 170 percent scram trip setting, none of t.he abnormal operational transients analyzed violate the f uel Safety Limit and there is a substantial margin from fuel damage.
There-fore, the use of ilow-referenced scram trip provides even additional margin.
An increase in the AMtM scram trip setting would de-crease the margin present bef ore the fuel cladding integrity Saf ety Limit is reached.
The APRM scram trip setting was determined by an analysis of margins re-quired to provide a reasonable range for maneuvering during operation.
Reducing this operating margin would increase the f requency of spurious scrams which have an adverse effect on reactor safety because of the result-ing thermal stresses.
Thus, the APRM scram trip set-ting was selected because it provides adequate margin for the fuel cladding integrity Safety Limit yet allows an operating margin that reduces the possibility of l
unnecessary scrams.
lhe scram trip setting must be ad.iusted to ensure that the LHGR transient peak is not increased for any com-bination of maximum fraction of limiting power density (MFLPD) and reactor core thermal power.
The scram I
setting is ad. justed in accordance with the formula in Specification 2.1.a.l.a. When the MFLPD is greater than the fraction of rated power (FRP).
This adjustment may :
i
g he accennpl i shed by increasing the 'APRM qain, and thus rea1oc i mi the slope and inst ercept point of t.he flow-referenced Al'HM H)qh Ilux Noram Curve by the reciprocal of the APRM qain thonge.
Analyses of the limiting trdaisients show that no scram ddj ust ment is requt red i n ossure MceH above the safety limit when the trainsient is i rii t i d ted from t.he operat-ing MCPR 1imi1.
The licensee provided the f ollowing technical specification bases for rod-block t. rip settings:
Reactor power level may be varied by moving control rods or by va ry i sig the recirculation flow rate.
The APRM system provides a c osit rol rod block which is dependent on rec i rc ul at i on flow rate to limit rod withdrdwal; tinus prot ect inig against. a MCPR of less than the MCPR fuel cladding integrity saf et y limit.
The flow variable trip setting provides substantial margin f rom iuel datinage, essumitig a steady state operation a t.
the trip setting, over the entire recirculation flow rosige.
The margin to the Saiety Limit increases as the flow decredses Ior the speCiiled ITip settin9 versus flow relationship; theref ore the worst case MCPR which could occur during steddy state operation is at 108% of roted t hensia l power because of the APRM rod block trip setting.
The act ual power dist ribut ion in the core is established by specified cont rol rod sequences and is monitored continunusly by the in-core i PRM system.
As with the APRM scram t rip setting, the APHM rod block trip setting is odjusted downward if the maximum frac-tion nf limit tnq powrr density exceeds the traction of rated power; thus preserving the APRM rod block safety margin.
As with the scroin setting, this may be accom-plished by odjusting the APHM gain.
The licensee indicated in ref erence 6 that CNS Procedure 10.1 entitled "APRM Calibration" was recently modified to include a provision for APRM gain od,j ustment to account for the difference between ef fective drive flow for single-loop and two-loop operation.
This modification involves adding the tenti 0.666W to 1.he APRM readings.
After completion of the APRM adjustment, t.he results are reviewed by the Shift Supervisor and the CNS Engineering Department.
This procedure ensures the necessary adjustments are performed properly.
Because sustained single-reci rcul ati on-loop operation is a rare event and not a planned mode of operation, we find 1.he above manual trip point setting procedures to be in accordance with the requirements of Section B.2 of BTp E1CSB-1? and accept-able for an interim period up to the next plant refueling.
However, in order to satisfy all review criteria, we recommend that the s i ngl e-recirculation-loop APRM and RBH trip set. points be made automatic or hard-wired (switch-selectable) f rom the cont.rol panel.
The changes should be
-3 l
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k.
r made at the first opportunity arvi installtd no later than the next CNS I
refueling out aqe.
't he'.e hardwire nwidificat. inns must. be submitted to the staf f f or review prior to iris t.a l l a t i on.
l The GT-Nf D report NfDti-74758 safety analyses were performed l
cssuming the recirculation equali7er valves were closed.
Further, the report indicated that the discharge valve in the idle recirculation loop is normally closed.
However, it its closure is prevented, the suction valve in the loop should he closed to prevent the loss of Low Pressure Coolant i
Injection (I pCI) flow nut of a post ul ated break in the idle loop suction line.
We reconenend that the licensee revise the proposed technical speci-fications to include t he requirement f or proper valve alignment and tagging prior to consnencement to single-recirculat.iun-loop operation.
The St ability Ariolysis section of NTDU-24?b8 indicrtes that the least stable power / flow condit. inns attainable under normal conditions occur at natural circulation with the control rods set for rated power and flow.
This condition may be reached following the t.ri p of both recirculation pumps.
One pump runniing at missimum speed 5 more stable than operating wit.h natural flow only, but is less stable than operating with both pumps operating at nit tiisitum speed.
Under single-recirculation-loop operation, the flow control shnuld be in master manual, since control oscillations may occur in the recirculat. inn flow cont rol system under these conditions.
We reconnend that the licensee revise the proposed technical specifications to include the requirement of master manual control of recirculation flow by the operator, as opposed to autoinatic control during single-recirculation-loou operation.
liecause of the diff erent flow pattern during single-recircula-tion-loop operation, a insuber of indicat.i ons in the cont rol room will change, such as individual jet-pump flow and total sunned core flow.
Some indications will be only slightly less than accurate, but some others will be erroneous.
All anomalous control room indications must be corrected or warning-tagged f ur the duration of the single-recirculation-loop operation, as required by section 4.?O of 1000 Std-279-1971.
III.
CONCLUSIONS Based on our review of the information and documents provided by the licensee, we conclude that t.he more conservative setpoints for'the APRM and RBM will be properly adjusted to protect the reactor for single-recirc-ulation-1 cop operat.1on.
The current manual method of setting the ApRM and RBM trip points is acceptable for an interim period of up to the next pl ant refueling outage.
In order to satisfy the requirement.s of the' review criteria, it will be necessary
- t. hat. these trip point. settings he made automatic or hardwired (switch-selectable) trinn the control panel.
The hardwire modifi-cations will requier statt review prior to installation.
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s In order to prevent the potential -loss of 1.pCl, we recommend that
/
the licensee revise the proposed technical specifications to include the require nents of proper valve aligivnent anni tagging prior to consnencewnt of single-recirculation-loop operation.
In order to achieve st.ahl e recirculat. ion flow control during single-rect reulat t osi-loop operat ion, we recosenend that. the licensee revise the proposed technical specificat. ions to include the requirement of mast.er manual control of recirculation t iow by the operator, as opposed to auto-matic control during single-recirculatiusi-loop operation.
All anoundlous cont.rol rmuu indications must. either be corrected for single-recirculation-loop operation or warning-tagged.
We conclude that. upon successful implementation of the above reconenended actions, the proposed licensee ansnendment for single-recircu-lation-loop operation at Cooper Nuclear Station is accept.able.
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REFERfNCES
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/
1.
Nebraska Public Power District letter (J.
M. Pilant) to NRC (1. A.
Ippolito), "Charige to Appesidix A Technical Specificatinns-Single Loop Operat ion," dated August 5,19H0.
2.
General flectric Csunpony, Nuclear Energy Division, " Cooper Nuclear Station Single Loop Operation,'" NLDU-242bH, by 1980.
3.
Nebraska Public Power District l et.ter (J.
M. Pflant) to NRC (D. B.
Vassallo). " Single Loop Operation-Response to NRC Questions," dated m y 6, 1982.
4.
Telephone conf erence call. NRC (R. Clark, B. Siegel); NPPD (J.
Weaver); EG&G San Manum (D. I andenbach), July 21. 1982.
5.
Telephone conierence cal 1, NRC (R. C1 ark, R.
Siegel, J.
T. Beard);
NPPD (J. Weaver); LGAG San Homon (D. I audenbach), July 22, 1982.
6.
Nebraska Public Power District letter (J.
M. Pilant) to NRC (D. B.
Vassallo), " Single 1. cop Dperation Revised Technical Speci fications,"
dated July 28, 1982.
7.
1EEE Std-279-1971:
Criteria for Protection Systems for Nuclear Power Generating Stations, dated 1971.
8.
NHC Branch lechnical Position E1CSH-12. " Protection System Trip Point Changes f or' Operat ion with kcactor Cool ant Pumps Out of Service,"
dated November 24, 1975.
9.
Code of F'-leral Regulations, Title 10, Part 50, Appendix A General Design Criteria ior Nuclear Power Plants, 1981.
10.
Project and Hudget Proposal for NRC Work, NRC Form 189.
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UhfTED STATES c
o NUCLEAR REGULATORY COMM!sslON t
g waSamorow, o. c. 2 ossa Q..ciR.f SEP y 0199g MEMORANDUM FOR: Themis P. Speis Assistant Director for Reactor Safety Division of Systems Integration FROM:
Brian W. Sheron, Chief, Reactor Systems Branch, DSI
SUBJECT:
BWR SINGLE LOOP-0PERATION - STATUS REPORT #2 Under multiplant action item E-04, there are 15 plants which involve single loop operation (SLO) the status of each plant is given below.
SER Licensing Plant Issued for 50%
Amendment Name Operation Issued Notes
- 1. Dresden-2 July 9, 1981 Yes
- 2. Dresden-3 July 9, 1981 Yds
- 3. Quad Cities-1 July 9, 1981 Yes
- 4. Quad Cities-2 July 9, 1981 Yes
- 5. Feach Bottom-1 May 15, 1981 Yes
- 6. Peach Bottom-2 May 15, 1981 Yes
- 7. Duane Arnold November 19, 1981 No 1
- 8. Cooper December 10, 1981 No 1
- 9. Pilgrim-1 December 15, 1981 No 1
- 10. Browns Ferry-1 August 16, 1982 No 1,2
- 11. Browns Ferry-2 August 16, 1982 No 1,2 1
- 12. Browns Ferry-3 August 16, 1982 No 1,2
- 13. Monticello September 10, 1982 No 1
- 14. Brunswick-1 Being Reviewed SER Due on November 1,1982
- 15. Brunswick-2 Being Reviewed. SER Due on November 1,1982 Notes 1.
Instrumention & Control review by EG&G, San Ramon Operations, CA is not completed. Licensing amendment will be issued after we get an acceptable TER from EG&G.
2.
CPB sections on thermal hydraulics and stability analysis are not included in the SER submitted for Browns Ferry Units 1,2,83.
As a follow up to the Browns Ferry-1 meeting with GE on single loop l
operation, questions were forwarded to T. Novak by memo' dated January l
15, 1982 from T. Speis for sending to GE and the utility group.
l Recently, approval was granted by GA0 for sending the questions to the l
licensees. Questions are expected to be sent to the individual I
If)f0*
eg g,d.1 re d _ i -
.i 4
. m " Y W T +^i" M
w~>**
Themis Speis.
SEP 2 01932 licensees during first week of October 1982.
Approval for single loop operation at a power greater than S0% can be granted only after staff concerns steming from Brnwns Ferry-Unit 1 single loop operation are satisfied.
Brian W. Sheron, Chief Reactor Systens Branch, DSI cc:
R. Mattson D. Vassalo R. Clark K. Eccelston BESiger J. VanVliet H. Nicolaras C. Berlinger RSB Section Leaders L. Phillips G. Schwenk 9
_ ___,