ML20027E396
| ML20027E396 | |
| Person / Time | |
|---|---|
| Site: | Haddam Neck File:Connecticut Yankee Atomic Power Co icon.png |
| Issue date: | 11/09/1982 |
| From: | Crutchfield D Office of Nuclear Reactor Regulation |
| To: | Counsil W CONNECTICUT YANKEE ATOMIC POWER CO. |
| References | |
| TASK-15-02, TASK-15-2, TASK-RR LSO5-82-11-023, LSO5-82-11-23, NUDOCS 8211150164 | |
| Download: ML20027E396 (18) | |
Text
i November 9, 1982 Docket No. 50-213 LS05-82 023 l
Mr. W. G. Counsil Vice President Nuclear Engineering and Operations Connecticut Yankee Atomic Power Company Post Office Box 270 Hartford, Connecticut 06101
Dear Mr. Counsil:
SUBJECT:
HADDAM NECK - SEP TOPIC XV-2, SPECTRUM OF STEAM SYSTEM PIPING FAILURES INSIDE AND OUTSIDE CONTAINMENT (SYSTEMSANDRADIOLOGICALCONSEQUENCES)
Enclosed is the staff's final evaluation of SEP Topic XV-2 for the system and radiological consequence aspects of this topic. The staff issued safety evaluations for this topic on May 20,1982 (systems) and July 15,1982 (radiological consequences) and has revised these evaluations based on comments provided in your letter of September 8, i
1982. The staff finds that both the system and radiological aspects of this topic meet current criteria provided that, for the radiological consequences, the Westinghouse Standard Technical Specifications requirements for primary and secondary coolant iodine specific activity as well as associated surveillance be adopted and implemented.
This evaluation will be a basic input to the integrated safety assess-ment for your facility unless you identify changes needed to reflect the as-built conditions at your facility. This assessment may be f
revised in the future if your facility design is changed or if NRC criteria relating to this subject is modified before the integrated
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assessment is completed.
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Sincerely,
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h h0M OkddO[3 Dennis M. Crutchfield, Chief
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PDR Operating Reactors Branch #5 Division of Licensing
Enclosure:
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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' 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
~ Haddam, Connecticut 06103 State of,C.onnecticut Office of" Policy and Management ATTN:
Under Secretary Energy Division 80 Vashington, 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 631 Park Avenue King of Prussia, Pennsylvania 19406
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HADDAM NECK NUCLEAR POWER STATION XV-2 SPECTRUll 0F STEAM SYSTEM PIPING FAILURES INSIDE/0UTSIDE CONTAINMENT (SYSTEMS)
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 trip, there is an increased possI6'ility 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 structures, systems and.
components of the facility with the objective of assessing the risk to public health and safety resulting from operation of the facility.
The steam.line break is one of the postulated accidents used to evaluate the adequacy of these structures, systems and components with respect to the public health and safety.
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. Section 50.36 of 10 CFR Part 50 requires the Technical Specifica-tions 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 Part 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 conjunction 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 stuckCFods, the capability to cool the core is maintained.
GDC 27, " Reactivity Limits," requires that the reactivity control systems be designed with appropriate limits on the potential
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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)'sufficiently disturb the core, its support structures, or other reactor pressure vessel internals to impair significantly the capability to cool the core.
GDC 27 specifically. requires that these postulated reactivity accidents include consideration of the steam line break accident.
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. GDC 31, " Fracture Prevention of Reactor Coolant Pressure Boundary,"
requires that the boundary be designed with sufficient margin to assure that when stressed under operating, maintenance, testing, and postulated accident conditions (1) the boundary behaves in a nonbrittle manner and (2) the probability of rapidly propagating fracture is minimized.
,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 limite'd to negligible amounts.
The system shou?6'have suitable redundancy and interconnections such that system function can be maintained assuming a single failure and assuming availability 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 l
SEP Topics III-5.A, and III-5.B consider the effects of the pipe break on safety related equipment.
SEP Topics VI-2.0 and VI-3 consider the ability of containment and the containment heat removal systems to mitigate the temperature /
pressure transient.
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. Other SEP topics address such items as ESF initiation, emergency 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 criteria specified in the Standard Review Plan are identified.
A separate evaluation is performed of the radiological consequences for conformance to 10 CFR Part 100.
V.
EVALUATION The atin steam system conducts steam in a 24-inch pipe from each of the four steam generators within the reactor containment through a nonreturn valve and a swing disc type trip valve into a common 36-inch' manifold.
From the manifold, the steam passes through two[
30-inch pipes to the two turbine stop trip valves.
Steam flow-meters are provided in the line from each steam generator, down-stream of the nonreturn' valve and trip valves.
The nonreturn valves prevent reverse flow of steam in case of a break upstream 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 ruputre 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
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. preventing a significant reactor coolant cooldown.
Since the reactor would trip on high steam flow before the core inlet 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 thiseccenario in reference 1.
The initial power increase will be terminated by an overpower trip.
A step decrease in total steam flow occurs at 9,5 seconds as a result of the trip.
The minimum DNBR is 1.65 and occurs 9.1 seconds after the break.
The maximum reactivity gain as a result of the cooldown is 1.01%
/1K.
This occurs at 100 seconds after the break.
Since the shut-down margin provided is calculated as 3.4% even with the highest worth rod fully out of the core, the reactor will remain sub-stantially subcritical a,t all times following the reactor trip.
Even with no boron injection by the Safety Injection System, the maximum reactivity gain would be 1.9%, which is still less 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.
Four 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' safety injection system actuated from low pressurizer pressure signals which, by addition of borated refueling water, ee-will prevent the reactor from returning to criticality after the reactor trip.
(3)
The steam line isolation trip valve circuit which is actuated,.
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upon coincidence of high steam flow in any two main steam lines.
Actuation of this circuit initiates a reactor and turbine trip and also closes all four steam isolation trip valves.
(4)
The main feedwater isolation circuit which is actuated upon high containment pressure.
Actuation of this circuit initiates closure of the feedwater isolation valves.
Depending on the location of the steam line rupture and the initial operating power level, a reactor and turbine trip will be initiated either by the isolation trip valve circuit, the over-power 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 initiation of auxiliary feedwater on return-to-power.
The analysis was performed with the RELAP4/M005 computer code.
Full feedwater flow was assumed until T goes below 535*F, then the feedwater ave control valves would ramp closed.
The main feedwater isolation system would act to close the feedwater isolation valves on high containment pressure.
The auxiliary feedwater (AFW) was assumed to be initiated at the time of the break.
A conservatively high*A7W 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 full power with and without offsite power available, and hot zero power with and without offsite power.
For all four cases, no return-to-power was predicted after scram.
The analyses show t
that for a steam line rupture during either four or three-loop l
operation there would be no core damage and DNBR would stay significantly above 1.3 (minimum value predicted was 1.62).
Also, for the maximum cooldown, no return-to-power would occur.
The licensee's steam line break analysis has assumed blow-down of only one steam generator, The remaining generators are assumed to be isolated by the MSIVS.
A failure of an MSIV to close in c.,
. 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 automatically trips the turbine (closes the turbine stop valves).
V.
CONCLUSION 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 af ter the incident.
There-fore,*%"b conclude that the Haddam Neck Plant is in conformance with Section 15.1.5 criteria of the Standard Review Plan.
VI.
REFERENCES 1.
Facility Description and Safety Analysis (FDSA).
2.
Plant Design Change #21 - October 1967, monthly report to AEC.
3.
W.
G.
Counsil letter to D. L. Ziemann, dated September 22, 1978, " Emergency Power Systems,"
4.
W. G. Counsil letter to D. L. Ziemann, dated September 29, 1978, " Proposed Revisions to Technical Specifications."
5.
W. G. Cou.nsil letter to D. L. Ziemann2 dated October 20, 1978, " Emergency Power Systems."
6.
W. G. Counsil letter to D. L. Ziemann, dated January 30, 1980, " Automatic Initiation of Auxiliary Feedwater."
7.
W. G. Counsil letter to D. M. Crutchfic1d, dated September 30, 1981, "SEP Section XV Topics, Design Basis Events."
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.i HADDAM NECK NUCLEAR POWER STATION XV-2 SPECTRUM OF STEAM SYSTEM PIPING FAILURES INSIDE AND OUTSIDE CONTAINMENT (RADIOLOGICAL CONSEQUENCES) 1.
INTRODUCTION Rupture of a steam line outside containment wiLL allow radioactivity contained in the primary and secondary coolant to escape to the environment.
SEP Topic XV-2 intended to review the radiological consequences of such failures.
The review wilL encompass those design features which limit the release of radioactivity in the released coolant and the amount of primary to secondary leakage.
II.
REVIEW CRITERIA S e c(j o_n 50.34 of 10 C FR Pa rt 50 requires that each appli cant for a construction permit or operating License provide an analysis and e va luation of the design and performance of structures, systems, and components of the f acility with the objective of assessing the risk to public health and safety resu lt ing from operation of.the f acilit y.
The steam Line break accident is one of the postulated i
accidents used to evaluate the adequacy of these structures, l
l systems, and components with respect to the public health and safety.
In addition, 10 C FR Pa rt 100.11 provides dose guidelines for reactor siting against which calculated accident dose consequences may be compared.
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2-III.
RELATED SAFETY TOPICS Topic II-2.C, " Atmospheric Transport and Diffusion Characteristics for Accident Analysis" provides the meteorological data used to evaluate the offsite doses.
Topic III-5.A, " Effects of Pipe Breaks
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on Structures, Systems and Components Inside Containment," and Topic III-5.B, " Pipe Break Outside Containment" will cover the dynamic effects of the postulated pipe failure inside and outside containment.
IV.
REVIEW GUIDELINES The review of the radiological consequences of these failures was conducted in accordance with Appendix A of Standard Review Plan Section 15.1.5, " Radiological Consequences of Main Steam Line Failure Outside Containment."
This review determines the effects ec-of fuel damage, iodine spiking and primary to secondary steam generator tube Leakage on Exclusion Area Boundary (EAB) and Low Population Zone (LPZ) Boundary doses after a steam Line break for--
two hours or the course of the accident, respectively,'then compares these doses to the dose guideline values.in 10 CFR Part 100.
The.
p la nt is considered adequately designed against steam line failures if the calculations show that the resultin,g of f site doses are less than a smalL fraction (10%) of the 10 CFR Part 100 exposure guidelines for the case of an accident-induced iodine spike and are within the exposure guidelines of 10 CFR Part 100 f or the case of a preaccident iodine spike or one rod held out of the Core.
. V.
EVALUATION In a letter to the NRC dated September 3 0, 19 81, t h e Licensee stated that because the accident produces no additional fuet f ailure and the radiological consequences from a main steam Line break accident are bounded by the radiological consequences of a steam generator tube rupture and a break in a smalL line carrying primary coolant outside containment, the radiological consequences for a main steamline break outside containment were not analyzed.
Therefore, the staff has performed an evaluation of the radiological consequences fotLowing a main steam Line break accident using what a re believed to be conservative assumptions as outlined below.
on-Releas e of a ctivit y following this accident is via the secondary system, but consists of activity originally contained in the secondary system aid that which passes f rom the primary system through steam gener stor tube Leaks.
The Licensee's curre7t technical specifications for primary coolant do not contain either an equilibrium or a maximum limit l
l for iodine activity, but rather contains an equilibrium Limit for total activity (alL isotopes with half-Lives greater than 1/2 hour).
The absence of any Limits on iodine activity in the primary and t
secondary coolants makes it impossible to evaluate the radiological consequences of this aciident.
Without such limits there can be no assurance that p la nt operation wouLd be restricted such that the exposu re guidelines of 10 CFR Part 100 wouLd not be exceeded in the event of a steamline break accident.
. As a starting point for the calculation the staff used the coolant (p ri ma ry and secondary) activity limits specified in the Westinghouse Standard Technical Specifications.
Two cases have been evaluated and both assume that no fuel failures occur as a result of the accident.
The first case (Case 1) assumes that the primary and secondary activities are at the maximum technical specification limit permitted in the Westinghouse Standard Technical Specifications, 60 pCi/ gram dose equivalent iodine-131 (DEI-131) a nd 0.1 pC i /gm DEI-131, respectively.
AlL the primary-to-secondary leakage is assumed to occur in the affected steam generator and is assumed to continue at.the technical specification rat'Eof 0.4 gallon per minute for a period of 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, after which the leakage is assumed to cease.
AlL the iodine in the primary-to-secondary leakage is assumed released to the environment.
In a Let t e r dat ed Sept embe r 8,1982, the licensee provided estimates of the steam released to the environment during the plant cooldown.
The staff used these steam release values and conservatively assumed that aLL the steam released had an iodine specific activity equal to 0.1 p C i / g r a m D EI-131.
Other appropriate values used in the staff calculations are presented in Table XV-2-1 of this evaluation.
Using these conservative assumptions, and the atmospheric dispersion factors generated in SEP Topic II-2.C, the resultant offsite radiological consequences are less than the guideline values of 10 CFR Part 100.
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1 j In the second case (Case 2), the staf f evaluated the radiological consequences of a main stealine break event assuming that the accident initiates an iodine spike which increases the specific activity of iodine in the primary coolant at the rate of 30 uci/
gram DEI-131 per hour.
The spike is assumed to increase the primary coolant activity over a period of 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> after which the coolant activity is assumed to remain constant at the 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> value.,For this case the staff, assumed that the primary coolant has an initial specific activity of 1.0 uCi/ gram DEI-131.
Using the conservative assumptions presented in Table XV-2-1 of this evaluation, the calculated offsite radiological consequences are Le(Y:than a smalL fraction of the 10 CFR Part 100 guideline values.
The staff's estimates of the offsite radiological consequences i
at the Exclusion Area and Low Population Zone boundaries are presented in Table XV-2-2.
The whole body doses are smalL, and because they do not approach the acceptance criteria, they are not presented in the table.
VI.
CONCLUSION Based upon the staff's evaluation, the potential radiological consequences following a main steamline break at the Exclusion Area and Low Population Zone Boundaries are less than the acceptance criteria given in SRP Section 15.1.5 Appendix A and~ the guideline values of 10 CFR Part 100.
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% However, the staff's calculations were based upon specific li mi t i ng values for iodine activity in the prim'ary and secondary coolants.
In order to assure that the staff findings presented above are app rop ri at e at the Haddam Neck' facility, the staff recommends that the Licensee i mp le ment the Westinghouse Standard Technica L specification (STS) Limits for specific activity in the primary and secondary coolant and the accompanying surveillance requirements.
The staff analysis is based on the Licensee's statement in the-September 30, 1981 Letter that no additional fuel 'f ailures occur as a result of this accident.
With the imp lement ation of the Westinghouse STS f or primary and w> -
secondary coolant activities presented above, the staff concludes that the plant is adequately designed to mitigate the radiological consequences of a main steamline break accident.
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TABLE XV-2-1 ASSUMPTIONS MADE IN ANALYSIS OF THE RADIOLOGICAL CONSEQUENCES OF THE MAIN STEAMLINE BREAK OUTSIDE CONTAINMENT ACCIDENT 1.
Reactor power = 1825 MWth 2.
Loss of offsite power following the accident 3.
Iodine decontamination f actor of 1 between water and steam for primary to secondary leakage (s team generator is dry).
a 4.
No additional fuel f ailure is assumed (Licensee estimate).
5.
P r,i ma ry to seconda ry leak, rate of 0.4 gpm to the affected steam generator for a period of 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.
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6.
Primary coolant specific activity prior to the accident
- Case 1:
6 0 u C i /g m D EI-131.
dcCase 2:
1.0 uCi/gm DEI-131.
7.
Secondary coolant specific activity prior to the accident of 0.1 uti/ gram of DEI-131.
8.
Iodine spiking factor in the primary coolant of 30 uti/ gram-hrs DEI-131 for the first four hours (C,a s e 2, a n a c c i de nt induced iodine spike).
I 9.
Atmospheric dispersion f actors (sec/ cubic meter) f rom SEP Topic II-2.C:
Exclusion Area, Boundary (0-2 hour) = 8.4 E-4 Low Population Zone Boundary (0-8 hours) = 5.4 E-5 10.
Total secondary coolant assumed to be released to the environment:
0 - 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> = 375,000 lbs 0 - 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> = 860,000 lbs e
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.o TABLE XV-2-2 THYROID DOSES FOLLOWING A MAIN STEAMLINE BREAK ACCIDENT Thyroid Doses (Rem)
Case Exclusion Area Low Population Boundary Zone Boundary Case 1 (PCA* = 60 uci/ gram DEI-131) 10.8 2.0 Case 2 (PCA* = 1 uCi/ gram DEI-131 9.0 2.4 with a concurrent iodine spike)
- PCA = Primary Coolant Activity t o.-
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