ML20053A615
| ML20053A615 | |
| Person / Time | |
|---|---|
| Site: | Haddam Neck File:Connecticut Yankee Atomic Power Co icon.png |
| Issue date: | 05/20/1982 |
| From: | Crutchfield D Office of Nuclear Reactor Regulation |
| To: | Counsil W CONNECTICUT YANKEE ATOMIC POWER CO. |
| References | |
| TASK-15-02, TASK-15-06, TASK-15-2, TASK-15-6, TASK-RR LSO5-82-05-043, LSO5-82-5-43, NUDOCS 8205270012 | |
| Download: ML20053A615 (18) | |
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i May 20,1982 Docket No. 50-213 LS05 05-043 Mr. W. G. Counsil Nuclear Engineer and Operations Connecticut Yankee Atomic Power Company Post Office Box 270 Hartford, Connecticut 06101
Dear Mr. Counsil:
SUBJECT:
HADDAM NECX - SEP TOPICS XV-2 SPECTRUM OF STEAM SYSTEM PIPING FAILURES INSIDE AND OtfrSIDE CONTAINMENT (SYSTEMS)
AND XV-6 FEEDWATER SYSTEM PIPE BREAXS INSIDE AND OUTSIDE CONTAINMENT (PWR)
By letter dated September 30, 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 complete the systens review of these topics for the Inddam Neck pl ant.
Potential radiological consequences of steam system failures will be addressed in a separate evaluation.
i 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 completed.
Sincerely, Dennis M. Crutchfield, Chief Operating Reactors Branch #5 Division of Licensing
Enclosures:
As stated cc w/ enclosures:
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Mr. W. G. Counsil cc Day, Berry & Howard Counselors at Law One Constitution Plaza Hartford, Connecticut 06103 Superintendent Haddam Neck Plant RFD #1'
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Post Office Box 127E East Hampton, Connecticut 06424
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. Mr. Richard R. Laudenat.
Manager, Generation Facilities Licensing Northeast Utilities Service Company P. O. Box 270~
Hartford, Connecticut 06101 Board of Selectmen Town Hall
'Haddam, Connecticut 06103 State of Connecticut Office of Policy and Management ATIN:
Under Secretary Energy Division 80 Washington Street Hartford, Conndcticut 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 631 Park Avenue King of Prussia, Pennsylvania 19406
'5 HADDAM NECK SEP TOPIC XV-2
_ SPECTRUM OF STEAM SYSTEM PIPING FAILURES INSIDE/0UTSIDE CONTAINMENT I.
INTRODUCTION A steam line break in the secondary system results in an in-crease in steam flow which decreases during the accident as the steam pressure falls.
The energy removal from the reactor coolant system causes a reduction of coolant temperature and pressure.
In the presence of a negative moderator tempera-ture coefficient, the cooldown results in a reduction of core shutdown margin.
If the most reactive rod cluster control assembly (RCCA) is assumed stuck in its fully with-drawn position after reactor trips there is an increased possibility that the core will become critical and return to power.
II.
REVIEW CRITERIA Section 50.34 of 10 CFR Part 50 requires that each applicant for an operating license provide an analysis and evaluation of the design and performance of structuresi systems, and components of the facility with the objective of assessing l
the risk to public health and safety resulting from opera-l tion of the facility.
The steam line break is one of th'e i
j postulated accidents used to evaluate the adequacy of these I
structuress systemsr and components with respect to the j
public health and safety.
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k 5-Sec t ion 50.36 of 10 CFR Part 50 requires the Technical Specifications to include safety limits which protect the integrity of the physical barriers which guard against the uncontrolled release of radioactivity.
The General Design Criteria (Appendix A to 10 CFR Pcrt 50) establish minimum requirements for the principal design criteria for water-cooled reactors.
GDC 27, " Combined Reactivity Control System Capability,"
requires that the reactivity control systems, in conjunc-tion with poison addition by the emergency core cooling system, has the capability to reliably control reactivity changes to assure that under postulated accident conditions, and with appropriate margin for stuck rods, the capability to cool the core is ma'intained.
GDC 28, " Reactivity Limits," requires that the reactivity control systems be designed with appropriate limits on the potential 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 boundary greater than limited local yielding, nor (2) suf-ficiently disturb the corer its support structuress or other 0
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reactor pressure vessel internals to impair significantly the capability to cool the core.
GDC 28 specifically re-quires that these postulated reactivity accidents include consideration of the steam line break acrident.
" Fracture Prevention of Reactor Coolant Pressure GDC 31, Boundary," requires that the boundary be designed with suf-ficient margin to assure that when stressed under operatinge
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maintenancer testings and postulated accident conditions (1) the boundary behaves in a nonbrittle manner and (2) the probability of rapidly propagating fracture is minimized.
i GDC 35, " Emergency Core Cooling," requires that a system be provided to, provide abundant emergency core cooling whose function is to transfer heat from the core following a loss of coolant such that (1) fuel and clad damage that could interfere with continued effective core cooling is prevented and (2) clad metal water reaction is limited to negligible amounts.
The system should have suitable re-dundancy and interconnections such that system function can be maintained assuming a single failure and assuming avail-ability of only on site or only off-site power supplies.
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10 CFR Part 100.11 provides dose guidelines for reactor sit-ing against which calculated accident dose consequences may be compared.
III.
RELATED SAFETY _ TOPICS SEP Topics III-5.A and 111-5.8 consider the effects of the pipe break on safety related equipment.
SEP-Topics VI-2.0 and VI-3' consider the ability of contain-ment and the containment heat removal systems to mitigate the temperature / pressure t ransient.
Other SEP topics address such items as ESF initiatione emer-gency power supplies, and containment isolation.
IV.
REVIEW GUIDELINES The review was conducted in accordance with SRP 15.1.5.
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 required.
Deviations from the cri-teria specified in the Standard Review Pla'n are identified.
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A separate evaluation is performed of the radiological con-sequences for conformance to 10 CFR Part 100.
V.
EVALUATION The main steam system conducts steam in a 24-inch pipe from -
each of the four steam generators within the reactor contain-ment through a nonreturn valve and a swing disc' type trip valve into a common 36-inch manifold.
From the manifolds the
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s t'e'a m 'p a s s e s through two 30-inch pipes to the two turbine stop trip valves.
Steam flowmeters are provided in the line from each steem generator, downstream of the nonreturn valve and trip valves.
The nonreturn valves prevent reverse flow of steam in case of a break upstream of the nonreturn valves.
If a pipe rupture occurs downstream of the trip valve, the increase steam flow will cause these valves to trip closed, thus limiting the blowdown period to just when the valves are open.
The maximum size steam line rupture is a circumferential double-ended rupture of the 36-inch main steam header.
This rupture can be quickly isolated from the steam generators, by rapid closure of the four main steam isolation valves (MSIVS),,thus preventing a significant reactor coolant cooldown.
Since the reactor would trip on high steam flow before the core intet temperature drops, there is no reactor power increase and the DNBR remains acceptable.
Also, the closure of the MSIVs would
terminate the cooldown and no return-to power would occur.
The reactor pressure would not fall below the safety injection signal setpoint.
The break location resulting in the most severe transient is a break between a nonreturn valve and the associated steam generator.
The affected steam generator would continue to blowdown.
The nonreturn valve in the line will eliminate
- blowdown-from the other steam' generators.
The licensee analyzed this scenario in reference 1.
The initial power increase will be terminated by an overpower trip.
A step de-crease in total steam flow occurs at 9.5 seconds as a result of the trip.
The minimum DNBR i s 1.65 and oc cu r s 9.1 seconds after the break.
The maximum reactivity gain as a result of the cooldown i s 1.01 % AK.
This occurs at 100 seconds after the break.
Since the shutdown margin provided is calculated as 3.4% even with the highest worth rod fully out of the corer the reactor will remain substantially suberitical at all times following the reactor trip. Even with no boron injec-tion by the Safety Injection Systems the maximum reactivity
. gain would be 1.9%, which is still tess than the shutdown margin available in rods.
For both of the steam line ruptures discussed above, no fuel damage or DNB was predicted to occur.
The reactor would not experience return to power over the course of the accident.
Three independent protection systems are provided to prevent core damage after a steam line rupture.
These protection systems are:
(1 )
The overpower and variable low pressure trip that provide overpower-overtemperature protection which, by tripping the reactor, will maintain an adequate margin to DNB during the initial phase of the transient.
(2) The sa'.ety injection system actuated from low pressurizer pressure signals which, by addition of borated refueling water, will prevent the reactor from returning to critic:,!ity after the r.eactor trip.
(3)
The steam line isolation trip valve circuit which is actuated upon coincidence of high s' team flow in any two main steam lines.
Actuation of this circuit initi-ates a reactor and turbine trip and also closes att four steam isolation trip valves.
O.
Depending on the location of the steam line rupture and the initial operating power levetia reactor and turbine trip wiLL be initiated either by the isolation trip valve circuit, the overpower trip circuit, or the variable low pressure trip circuit.
Long-term decay heat removal is accomplished by manual control of the steam dump valves and feedwater.
The above cases were analyzed in Reference 1 and were later reanalyzed and discussed in references 2, 3, 4, 5, and 6.
Reference 6 was concerned with the effect of automatic initi-ation of auxiliary feedwater on return-to power.
The analysis was performed with the RELAP4/ MODS computer code.
The feed-water system was assumed to operate as it normally would, futt flow until T,y,,g'oes b'e t o w 535 F,
then the feedwater valves ramp closed.
,The auxiliary feedwater (AFW) was assumed initiated at time of break. A conservatively high AFW flow was used as well as failure of one HPSI and one charging pump. The highest worth control rod was assumed stuck in its worst possible position.
Four cases were analyzed, hot futt power with and without offsite power available, and hot zero power with and without offsite power.
For att four cases, no return-to-O y
-9 power was predicted after scram.
The analyses show that for a steam line rupture during either four or three-loop operation there would be no core damage and DNBR would stay
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significantly above 1.3 (minimum value predicted wa s 1.62).
Also, for the maximum cooldown, no return-to power would occur.
The licensee's steam line break analysis has assumec blow-down of only one steam
'enerator.
The remaining generators are assumed g
to be isolated by the MSIVS.
A failure of an MSIV to close in an intact generator is not assumed to negate isolation of that generator because credit is taken for the turbine control and stop valve closure downstream of the MSIVS.
Credit for the control and stop valves is recognized by the staff, in NUREG-0138, provided the closure signal is derived from the protection system.
Opening of the reactor trip breakers au'tomatically trips the turbine lcloses the i
turbine stop valves).
V.
CONCLllSION The staff reviews of the steam line break have concluded that neither DNB nor fuel damage will occur as a result of a steam line rupture incident, and that the reactor will be promptly tripped and will remain subcritical after the incident.
Thereforer we conclude that the Haddam Neck Plant is in conformance with Section 15.1. 5 c r i t e r i a of the Standard Review Plan.
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REFERENCES 1.
Facility Description and Safety Analysis CFDCA).
2.
Plant Design Change #21 - October 1967e monthly report to AEC.
3.
W.
G.
Counsit letter to D.
L.
Ziemann, dated September 22, 1978, " Emergency Power Systems."
4.
W.
G.
Counsit letter to D.
L.
Ziemann, Dated September 29, 1978, " Proposed Revisions to Technical Specifications."
5.
W.
G.
Counsil letter to D.
L.
Ziemanni dated October 20, 1978,
" Emergency Power Systems."
6.
W.
G.
Counsit letter to D.
L. Ziemanni dated January 30, 1980,
" Automatic Initiation of Auxiliary Feedwater."
7.
W.
G.
Counsil letter to D.
M.
Crutchfield, dated September 30, 1981, "SEP Se c t i on XV Topic s, D e s ign B a s i s Event s."
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SEP TOPIC XV-6 FEEDWATER SYSTEM PIPE BREAKS INSIDE AND OUTSIDE CONTAINMENT (PWR)
I.
INTRODUCTION A feedwater system pipe break would result in the loss of main feedwater supply to the affected steam generator (s).
The time to recovery of feedwater supply depends on the location of the break and on operator actions.
II.
REVIEW CRITERI A Section 50.34 of 10 CFR Part 50 requires that each applicant for a construction permit or operating license provide an analysis and evaluation of the design and performance of structuress systems, and components of the facility with the objective of assessing the risk to public health and safety resulting from operation of the facilityi 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 Specifications to include safety limits which protect the integrity of the physical barriers which guard against the uncontrolled release of radio-activity.
C-i i
-2_
The General Design Criteria (Appendix A to 10 CFR Part 50) establish minimum requirements for the principal design criteria for water cooled reactors.
GDC 27r " Combined Reactivity Control System Capability," requires that the reactivity control systems, in conjunction with poison addition by the ' emergency core cooling systems has the capability to reliably control reactivity changes to assure that under postulated accident conditions, and with appropriate margin for stuck rodse the capability to cool the core is maintained.
GDC 28r " Reactivity Limits," requires that the reactivity control
~
systems be designed with appropriate limits on the potential.
am?unt and rate of reactivity increase to ensure that the effects of postulated reactivity a c c i,d e n t s can neither (1) result in damage to the reactor coolant pressure boundary greater than limited local yielding, nor (2) sufficiently disturb the corer its support structures or other reactor pressure vessel internals to impair significantly the capability to cool the core.
GDC 31, " Fracture Prevention of Reactor Coolant Pressure Boundary,"
requires that the uoundary be designed with sufficient margin to assure that when stressed under operatings maintenancer testing
and postulated accident conditions, (1) the boundary behaves in a nonbrittle manner, and (2) the probability of rapidly propagating fractures is minimized.
1 GDC 35, " Emergency Core Cooling," requires that a system be provided to provide abundant emergency core cooling whose function is to transfer heat from the core following a loss of coolant such that
( 1. ) fuel and clad damage that could interfere with continued effective core cooling is prevented, and (2) clad metal water reaction is limited to negligible amounts.
The system should have suitable redundancy and interconnections such that system function can be maintained assuming a single failure and assuming availability 1
of only on-site or only off site power supplies.
10 CFR Part 100.11 provides dose guidelines for reactor siting against which calculated accident dose consequences may be compared.
III.
RELATED SAFETY TOPICS SEP Topics III-5.A, " Effects of Pipe Break on Structures, Systems and Components Inside containment," and III-5.B, " Pipe Break Outsi6 Containment," consider the dynamic effects (pipe whip, j et
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impingemente adverse environment) on safety-related equipment.
Other SEP topics address such items as ESF initiation, the auxiliary feedwater system capacity, and containment isolation.
IV.
REVIEW GUIDELINES The review is conducted in accordance with SRP 15.2.8.
The evaluation includes review of the analysis of the event.
Identification of the features in the plant that mitigate the consequences of the event as well a,s the ability of these systems to function is required.
Deviations from the criteria specified in the SRP are identified.
V.
EVALUATION The feedwater pipe break event was analyzed with the 12F version of RETRAN (Ref. 1) code and the assumption of 103% of rated power.
The operator was assumed.to isolate the break within 10 minutes.
Main feedwater was assumed to be unavailable. The auxiliary feedwater system would be automatically initiated,' and following isolation of the break at 10 minutes, would be delivering feedwater to the unaffected steam generators at 101/2 minutes after the accident.
A feedwater line break between the feedwater check valve inside containment and the steam generator would result in the blowdown of that steam ge. iera tor.
This results in an initial cooldown of the primary
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system.
Main and auxiliary feedwater to all steam generators would be lost until the break is isolated.
This event is less severe from an overpressurization standpoint than the feedwater pipe break upstream of the feedwater check valve due to the initial reactor coolant system cooldown.
The cooldown transient of, this event is bounded by the steam line break event (Ref. 2).
The steam generator would not blowdown if the feedwater line break occurs upstream of the feedwater check valve.
Main feedwater and auxiliary feedwater would be lost until the break is isolated.
Because of the relatively.
Long steam generator dry out timer (approximately 30 minutes) (Ref. 1)r the operator would have sufficient time to isolate the break and ensure feedwater delivery to the unaffected steam generators. Since the reactor and the turbine have already trippede the only energy that needs to be dissipated is the core decay heat.
One of the two' steam driven auxiliary feedwater pumps (450 gpm e o'c h ) would have sufficient flow capacity to remove the decay heat and prevent the primary system from exceeding 110% of design pressure.
V.
CONCLUSIONS As part cf the SEP review for Haddam Neck Plant, the staff has evaluated the licensee's analysis of the feedwater pipe break event. The results indicate that the operator would have sufficient time to isolate the break to divert the auxiliary feedwater flow to the unaffected steam generators (Fef.1 ).
The maximum reactor coolant system pressure remains l
below 110% of the design pressure. We, therefore, find the results of the analysis for the feedwater pipe break event acceptable.
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6-VII.
REFERENCES
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1.
Letter from W. G. Counsil to J. M. Hendrie, dated November 30, 1979.
2.
Letter from W. G. Counsil to D. M. Crutchfield, dated September 30, 1981.
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