ML20052A337
| ML20052A337 | |
| Person / Time | |
|---|---|
| Site: | Haddam Neck File:Connecticut Yankee Atomic Power Co icon.png |
| Issue date: | 04/26/1982 |
| From: | Crutchfield D Office of Nuclear Reactor Regulation |
| To: | Counsil W CONNECTICUT YANKEE ATOMIC POWER CO. |
| References | |
| TASK-09-04, TASK-15-09, TASK-15-14, TASK-15-9, TASK-9-4, TASK-RR LSO5-82-04-071, LSO5-82-4-71, NUDOCS 8204280227 | |
| Download: ML20052A337 (13) | |
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q April 26,1982 Docket No. 50-213 LS05-82 071
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Mr. W. G. Counsil, Vice President
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Dear Mr. Coons 11:
SUBJECT:
HADDAM NECK - SEP TOPIC XV-9, STARTUP 0F AN INACTIVE LOOP; XV-14 INADVERTENT OPERATION OF ECCS AND CHEMICAL AND VOLUME CONTROL SYSTEM MALFUNCTION THAT INCREASES REACTOR COOLANT INVENTORY AND IX-4, BORON ADDITION SYSTEM By letters dated September 30, 1981 and October 2, 1981, you submitted safety assessment reports for the above topics. The staff has reviewed these assessments and our conclusions are presented in the enclosed safety evaluation reports, which completer the review of these topics for the Haddam Neck plant.
These evaluations will be a basic input to the integrated assessment for your facility. The evaluations may be revised in the future if your facility design is changed or if NRC criteria relating to these topics are modified before the integrated assessment is cogleted.
Sincently, Dennis M. Crutchfield, Chief fW 3
Operating Reactors Branch No. 5 Division of Licensing
Enclosures:
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Docket No. 50-213 Mr. W. G. Counsil Revised 3/30/82 cc Day, Berry & Howard Counselors at Law One Constitution Plaza Hartford, Connecticut 06103 Superintendent Haddam Neck Plant RFD #1~-
Post Office Box 127E East Hampton, Connecticut 06424 Mr. Richard R. Laudenat Manager, Generation Facilities Licensing Northeast Utilities Service Company P. O. Box 270 Hartford, Connecticut 06101 Board of Selectmen Town Hall
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'Haddam, Connecticut 06103 State of Connecticut 0Ffice of Policy and Management ATTN:
Under Secretary Energy Division 80 Washington Street Hartford, Connecticut 06115
'U. S. Environmental Protection Agency Region I Office.
ATTN:
Regional Radiation Representative JFK Federal Building Boston, Massachusetts 02203 Resident Inspector Haddam Neck Nuclear Power Station c/o 'U. 'S..NRC East Haddam Post Office East Haddam, Connecticut 06423 Ronald C. Haynes, Regional Administrator Nuclear Regulatory Commission, Region I l
631 Park Avenue King of Prussia, Pennsylvania 19406 C
6 HADDAM NECK SEP TOPIC IX-4 BORON ADDITION SYSTEM I.
INTRODUCTION
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following a LOCA, boric acid solution is injected into the reactor core from the refueling water storage tank via the High Pressure Safety Injection (HPSI) pumps, the Low Pressure Safety Injection (LPSI) pumps and the charging pumps (if offsite power is available).
During the recirculation phase, suction is taken from the containment building sump through the residual heat removal (RHR) pumps and heat exchanger and back into the reactor vessel. A portion of the coolant is converted to steam by the decay heat of the fuel.
Because the steam flowing out of the break has few impurities, the boric acid concentration of the remaining coolant is increased.
Unless a dilution flow is provided to reduce the boric acid concentration, it is possible to eventually precipitate solid boric acid which could clog coolant passages.
II.
REVIEW CRITERIA AND GUIDELINES The guidelines for this review are contained in Reference 1, which is a memo describing the methods used to review boric acid buildup post-LOCA long-term cooling.
There is no SRP section covering this topic.
This topic stems from the long-term cooling requirements of 10'CFR 50,46.
Although the Haddam Neck plant is not bound by the provisions of 10 CFR 50.46 since the fuel in the core utilizes stainless steel clad, similar long-term cooling requirements exist in the Interim Acceptance Criteria which do apply to Haddam Neck.
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III.
EVALUATION To prevent the buildup of boron in the core du-ing long-term recirculation, Haddam Neck utilizes simultaneous cold leg injec-tion with the charging pumps and upper plenum injection with the RHR pumps.
This method will prevent boron precipitation for both hot and cold leg breaks as discussed in References 2 and 3.
' The operator is instructed to open both paths from the RHR dis-1 charge.to the charging pump suction during long term recirculation.
All valves inside containment are locked in position so that flooding is not a problem.
Two paths are available so that a single failure will not prevent establishing a sufficient diluting flow.
IV.
CONCLUSION The staff has reviewed the methods and equipment used to control post LOCA boron concentrations and finds them acceptable.
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e REFERENCES l.-
Memorandum for T.
M.
Novak, Chief, Reactor Systems Branch, from K.
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Parcze wski, Reactor Safety Branch, January 21, 1976.
a 2.
Letter to D.
M. Crutchfield, NRC, from W.
G.
Counsil, CYAPCO, "Haddam Neck Plant, SEP Topic IX-4, Boron Addition System",
dated October 2, 1981.
3.
Letter to R.
A.
Purple, NRC, from D.
C.
Switzer, CYAPCO, "Haddam Neck Plant, ECCS Long Term Cooling Capability", dated September 29, 1975.
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INTRODUCTION The startup of an inactive coolant loop at an incorrect tem-perature is examined to assure that the consequences are acceptable.
The guidance for review of this topic is pro-
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vided by SRP sections 15.4.4 and 15.4.5.
The concern is that the influx of cooler water will cause an increase in core power (due to a negative moderator reactivity coeffi-cient) which will reduce thermal margins.
The calculated minimum departure from nucleate boi-ling ratio (DNBR) is compared to the acceptable minimum DNBR limit to demonstrate that fuel failures will not occur.
II.
REVIEW CRITERIA Section 50.34 of 10 CFR Part 50 requires that each applicant for a construction permit or operating license provide an ana-lysis and evaluation of the design and performance of struct-ures, systems, and components of the facility with the object-ive of assessing the risk to public health and safety resulting from operation of the facility, including determination of.
the margins of safety during normal operations and transient conditions anticipated during the life of the facility.
Section 50.36 of 10 CFR Part 50 requires the Technical Speci-fications to include safety limits which protect the integrity s%
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lof the physica1' barriers which guard against the uncontrolled release of radioactivity.
s The General' Design Criteria (Appendix A to 10 CFR Part 50) set forth the criteria for the design of water-cooled reac-tors.
GDC 10 " Reactor Design" requires that the core and associated cooling, control, and protection systems be designed with appropriate margin to assure that specified acceptable fuel design limits are not exceeded during normal operation, in-cluding the effects of anticipated operational occurrences.
GDC 15 " Reactor Coolant System Design" requires that the reactor coolant and associated protection systems be de-signed with sufficie~nt margin to assure that the design con-ditions of the reactor coolant pressure boundary are not exceeded during normal operation, including the effects of anticipated operational occurrences.
GDC 20 " Protection System Functions" requires that the pro-tection system be designed to initiate aut6catically-the to assure that spe-operation of reactivity control systems cified acceptable fuel design limits are not exceeded as a result of anticipated operational occurrences.
GDC 26 " Reactivity Control System Redundancy and Capability" requires that the reactivity control system be capable of reliably controlling reactivity changes to assure that under
conditions of normal operation, including anticipated opera-tional occurrences, and with appropriate margin for malfunc-lions such as stuck rods, specified acceptable fuel design limits are not exceeded.
GDC 28 " Reactivity Limits" requires that the reactivity con-trol systems be designed with appropriate limits on the po-tential amount and rate of reactivity, increase to ensure that the effects of postulated reactivity accidents can neither (1) result in damage to the reactor coolant pressure bound-ary greater than limited local yielding, nor (2) sufficient-ly disturb the core, its support structures, or other reactor pressure vessel interhals to impair significantly the capa-bility to cool the core.
III.
RELATED SAFETY TOPICS Various other SEP topics evaluate such items as the reactor protection system.
The effects of single failures on safe shutdown capability are considered under Topic VII-3.
IV.
REVIEW GUIDELINES The ' review is conducted in accordance with SRP 15.4.4 and 15.4.5.
The evaluation includes review of the analysis for the event.
and identification of the features in the plant that mitigate the consequences of the event as well as the ability of these systems to function as requir6d.
The extent to which. operator action is required is also evaluated.
Deviations from the cri-teria specified in the Standard Review Plan are ident.ified.
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V.
EVALUATION The startup'of an inactive loop was evaluated by the licensee in references I and 2.
The assumptions used in the analysis were:
1.
Initial power is 78% of full power, which is above the 65% limit for 3 loop o'peration.
2.
Water in isolated loop is 30 F lower than in other cold leg loops (interlock prevents opening isolated loop when temperature differs 20 F).
3.
Overpower trip is delayed.
Most negative moderator temperature coefficient and 4.
minimum doppler coefficient are used.
5.
Full flow in idle loop is reached in 1/2 the time it takes an isolation valve to open.
The method of analysis was with an analog simulation.
Two cases were considered; one with complete flow mixin'g between loops and the other with no flow mixing between loops.
Results of the analysis show a pressure increase of about 10 psi, and a peak nuclear flux of 94% of rated peak hot The minimum DNBR calculated was,1.94, which spot heat flux.
is above the acceptance criteria limit of 1.3.
VI.
CONCLUSION The staff has reviewed the Haddam' Neck Plant sutnittal on SEP Topic XV-9, Startup of an Inactive Loop. The results of the analysis indicate that the Haddam Neck Plant is in conformance with the SRP section 15.4.4 and 15.4.5 acceptance criteria and is acceptable.
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- 1. - Haddam Neck' Final Design Safety-Asalysis, Section'
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Haddam Neck Plant Design Change #21, October 1967.
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HADDAti NECK SEP Topic XV-14 INADVERTENT OPERATION OF ECCS AND CVCS MALFUNCTION THAT INCREASES REACTOR COOLANT INVENTORY I.
INTRODUCTION
. An inadvertent safety injection or a malfunction in the pressurizer level controls can cause an increase in primary coolant inventory. The increase in inventory may result in power level increase and lead to fuel damage or overpressurization of the primary system.
If the transient is severe enough the reactor will trip from high pressurizer water level or high pressurizer pressure.
II. REVIEW CRITERIA Section 50.34 of 10CFR Part 50 requires that each applicant for a construction permit or operating license provide an analysis and evaluation of the design and performance of structures, systas, and components of the facility with the objective of assessing the risk to public health and safety resulting from operation of the facility, including determination of the margins of safety during normal operations and transient conditions anticipated during the life of the facility.
Section 50.36 of 10CFR Part 50 requires the Technical Specifications to include safety limits whic, protect the integrity of the physical barriers h
which guard against the uncontrolled release of radioactivity.
The General Design Criteria (Appendix A to 10 CFR Part 50) establish minimum requirments for the principal design criteria for water-cooled reactors.
GDC 10 " Reactor Design" requires that the core and associated coolant, control and protection systems be designed with appropriate margin to assure that specified acceptable fuel design limits are not exceeded during nomal operation, including the effects of anticipated operational occurrence.
GDC 15 " Reactor Coolant System Design" requires that the reactor coolant and associated protection systems be designed with sufficient margin to assure that the design conditions of the reactor coolant pressure boundary are not exceeded during normal operation, including the effects of anticipated opera.tional occurr ences.
GDC 26 " Reactivity Control System Redundancy and Capability" requires that the reactivity control systems be capable of reliably controlling reactivity changes' to assure that under conditions of normal operation, including anticipated operational occurrences, and with appropriate margin for malfunctions such as stuck rods, spec-ified acceptable fuel design limits are not exceeded.
III.
RELATED SAFETY TOPICS Inadvertent boron dilution is discussed in Topic XV-10.
Various other SEP topics evaluate such items as the reactor protection systen.
The effects of single failures on safe shu'tdown capability are considered under-Topic VII-3.
IV.
REVIEW GUIDELINES The review is conducted in accordance with SRP 15.5.1, 15.5.2.
The evaluation includes review of the analysis for the event and identification of the features in the plant that mitigate the consequences of the event as well as the ability of these systens to function as required.' The extent to which operator action is required is also evaluated. Deviations from the criteria specified in the Standard Review Plan are identified.
V.
EVALUATION The licensee has evaluated the effects of an inadvertent safety injection during power operation (Ref.1).
It has been determined that since the primary L
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coolant pressure (2000 psig) exceeds the shutoff head of the safety injection pumps (1400 psig) no safety injection flow will be delivered durin,g power operation.
The licensee has further evaluated malfunctions of the pressurizer level control system that.could result in starting of the backup charging pump and isolation of letdown.
In this event the licensee has determined that indications and alarms in the control room will alert the operator of the mismatch in letdown and chargir.g flow and low level in the volume control ta'n k.
If the malfunctions were not detected and corrected, it would take approximately 12 minutes to fill the pressurizer and until a high pressure reactor trip occurs.
IV.
CONCLUSION As part of the SEP review for Haddam Neck, we have concluded that the licensee evaluation of inadvertent operation of ECCS and CVCS malfunctions confinns that no adverse consequences will result from this event.
References 1.
Letter from W. G. Counsil, CYAPC, to D. M. Crutchfield, NRC,
Subject:
Haddam Neck Plant SEP Section XV Topics, Design Basis Events, dated September 30, 1981.
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