ML20204A278
| ML20204A278 | |
| Person / Time | |
|---|---|
| Site: | Hatch  |
| Issue date: | 04/28/1986 |
| From: | Rivenbark G Office of Nuclear Reactor Regulation |
| To: | Beckham J GEORGIA POWER CO. |
| Shared Package | |
| ML20204A279 | List: |
| References | |
| TAC-53327, TAC-53328, TAC-56049, TAC-59542, NUDOCS 8605120016 | |
| Download: ML20204A278 (31) | |
Text
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Docket No. 50-321/366
'Mr. J. T. Beckham, Jr.
Vice President, Nuclear Generation APR 28 E Georgia Power Company P.~0. Box 4545
-Atlanta, Georgia 30302
Dear Mr. Beckham:
We are reviewing the request to revise the Hatch Unit 1 and 2 Technical Specifications to reflect the addition of the closure of containment purge and vent valves on a high containment radiation signal as submitted by your letters dated September 5, 1984, August 20, 1985 and January 7, 1986.
In order to complete this review, we need the additional information described in the enclosure. You are requested to provide a written response, submitting the requested information, within 60 days of receipt of this letter.
The reporting and recordkeeping requirements of this letter affect fewer than ten respondents; therefore, OMB clearance is not required under P.L.96-511.
Sincerely, r
Cricinal 81 eod byj G
George Rivenbark, Project Manager BWR Project Directorate #2 Division of BWR Licensing Office of Nuclear Reactor Regulation
Enclosure:
Request for Additional Information cc: See next page
. DISTRIBUTION EDocket. File "
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Local PDR Gray File DMuller/GRivenbark OELD EJordan 8605120016 860428 BGrimes
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fDR ADOCK 05000321 JPartlow PDR SNorris ACRS(10) 0FC : DBL:PD#2
- DBL JDr2 DBL:PD#2 J._____:____________:________....:____________: ___________:___... ____
_____:___________:....________:.g________:____..______:_....._____:...._______
NAME :SN ris:
- GR nbark D
y DATE :4/Ah86:
- 4/d/86 4/g86 0FFICIAL RECORD COPY
g Nr. J. T. Beckham, Jr.
Edwin I. Hatch Nuclear Plant, Georgia Power Company Units Nos. I and 2 cc:
Bruce W. Churchill Esquire Shaw,Pittman,Potis&Trowbridge 1800 M Street, N.W.
Washington, D.C.
20036 Mr. L. T. Gucwa Engineering Department Georgia Power Company P. O. Box 4545 Atlanta, Georgia 30302 Mr. M. C. Nix, Jr., General Manager Edwin I. Hatch Nuclear Plant Georgia Power Company P. O. Box 442 Baxley, Georgia 31513 Mr. Louis B. Long Southern Company Services Inc.
P. O. Box 2625 Birmingham, Alabama 35202 Resident inspector U.S. Nuclear Regulatory Comission Route 1, P. O. Box 279 Baxley, Georgia 31513 Regional Administrator, Region II U.S. Nuclear Regulatory Comission.
l 101 Marietta Street, Suite 3100 l
Atlanta, Georiga 30303 Mr. Charles 4. Badger Office of Planning and Budget l
Room 610 l
270 Washington Street, S.W.
Atlanta, Georgia 30334 i
Mr. J. Leonard Ledbetter, Comissioner Department of Natural Resources l
270 Washington Street, N.W.
Atlanta, Georgia 30334 Chairman Appling County Comissioners County Courthouse Baxley, Georgia 31513 l
f Hatch Unit 1 and 2 Vent and Purge Line Isolation on High Radiation Request for Additional Information 1.
For all primary containment vent and purge lines, please provide the following information:
a) inside diameter of the line; b) expected flow rates (maximum and design) when open to the environment, c) frequency and duration of the line being open to the environment, and d) length of seismic Category I piping, including the isolation valves, outside primary containment (if diameter changes, please provide the lengths of each section).
2.
What are the restrictions, if any, on the number of lines being simultaneously open to the environment? Of these lines, identify those which would provide a leakage pathway from the containc.ent to the environment.
(Include inlet as well as exhaust lines and use the assumption that the drywell is at 2 psig.)
3.
In an April 10, 1984 letter from L. T. Gucwa of Georgia Power Company to J. F. Stolz of the NRC, it was stated that the high radiation isolation signal will originate from the high-range containment radiation monitors which were installed in response to NUREG-0737 Item II.F.1(3).
Please provide the following information:
a)
Where are the two detectors specifically located in the drywell?
b)
Do these monitors meet the requirements of Table II.F.1-3 in NUREG-0737? If not, please describe the differences, c)
What activity levels (in curies per cubic feet) in the effluent leaking out the open vent and purge lines correspond to the isolation signal setpoint of 138 roentgens per hour after taking into account instrument accuracy, calibration errors, and drift of the setpoint that could occur during the interval between calibrations?
Please provide the assumptions used in the exposure dose rate calculations.
d)
What are the normal radiation levels in the drywell? And what is the highest level expected in normal operation?
e)
What is the maximum time for closure of the vent and purge isolation valves on a high containment radiation signal? Include the time to reach the isolation signal setpoint, instrument delay times, and time for the valve to be in the fully closed position after actuator power has reached the operator assembly.
[
i.
4.
Please provide an analysis of the radiological consequences of a small line failure (e.g., instrument lines and sample lines) inside containment assuming direct flow of flashed coolant to the environment through an open vent or purge line with appropriate credit taken for mixing and dilution. The procedures and assumptions outlined in Standard Review Plan Section 15.6.2 (Radiological Consequences of the Failure of Small Lines Carrying Primary coolant Outside Containment) should be used in the analysis. Alternatively, please provide an analysis which demonstrates based on geometric configuration, expected flow rates, or other considerations that the small line failure case is bounded by the accident scenario outlined in the June 14, 1982 letter from T. J. Dente of the BWR Owner's Group to D. G. Eisenhut of the NRC (copy attached) which assumed the drywell to be at 2 psig.
5.
In the request to change the Technical Specifications for Hatch Unit 1 (transmitted in September 5, 1984, August 20, 1985 and January 7, 1986 letters from J. T. Beckham, Jr., of the Georgia Power Company to J. F.
Stolz of the NRC), proposed revisions to Unit 1 Tables 3.2-1 and 4.2-1 and Unit 2 Tables 3.3.2-1 and 3.3.2-2 are presented. Unit 1 Table 3.7-1 and Unit 2 Table 3.6.3-1, " Primary Containment Isolation Valves," also need to be revised. Please submit revised Tables which specifically show which isolation valves will close on a high containment radiation signal.
6.
Specifically, will the isolation valves (T48-F103) and 2748-F103 on the drywell and suppression chamber nitrogen supply lines for Unit 1 and Unit 2 respectively automatically close on a high containment radiation signal? If not, please provide the justification for not isolating this valve on high containment radiation.
7.
Your August 20, 1985 submittal indicates that there will be two operable trip systems with one operable channel per trip system for vent and purge line isolation on high containment radiation (see note b on Table 3.2-1) whenever primary containment integrity is required. Your April 10, 1984 submittal states that the Technical Specification will be revised to require that one radiation monitor which initiates high radiation isolation be operable at all times except cold shutdowns and refueling outages. Please resolve this apparent discrepancy.
,4 b OUJNERS' GflOUP u o... _
p.O. Sox 270 e Hortford. Connecticut 06101 * (203) 666 6911 X 5489 June 14,1982 BWROG-8222 U. S. Nuclear Regulatory Commission Division of Licensing Office of Nuclear Reactor Regulation Washington, D.C.
20555 Attention:
Darrell G. Eisenhut, Director Gentlemen:
SUBJECT:
SUPPLEMENT TO BWR OWNERS GROUP EVALUATION OF NUREG-0737 ITEM II.E.4.2(7)
References:
1)
Letter from T. J. Dente (BWR Owners Group) to D. G.
i Eisenhut (NRC) titled "BWR Owners Group Evaluation of NUREG-0737 Item II.E.4.2(7)," dated June 29, 1981 2)
Letter from D. G. Eisenhut (NRC) to T. J. Dante (BWROG) titled "NUREG-0737 Item II.E.4.2(7): Containment Isolation Dependability - Isolation on High Radiation,"
dated October 14, 1981 3)
Minutes of Meeting between NRC Staff and GE/BWROG on November 19, 1981 4)
Telecon between F. Hayes (GE) and D. Verre111 (NRC) on February 18, 1982 Reference 1 transmitted the results of the BWR Owners Group initial evaluation of NUREG-0737. Item II.E.4.2, Part 7.
That evaluation concluded that automatic isolation of the containment vent and purge valves on high containment radiation is not necessary for Mark I and II plants. The basis for that conclusion is that:
(1) the vent and purge valves are normally closed, (2) there already exist separate and diverse signals for automatic closure of those valves, (3) there are various signals which will alert the operator to manually close the valves, and (4) the radio-logical consequence from a break sufficiently small so as not to automatically isolate these valves is acceptably low.
The NRC rejected the Owners Group position in Reference 2.
In response to the NRC rejection, the BWR Owners Group met with the NRC (Reference 3) in order to obtain a clarification of the principal NRC concerns that prompted this requirement as well as a clarification of the requirement itself. A presentation was given to the Staff by the BWR Owners Group in which it was shown that the benefit of adding an automatic radiation signal is negligible due to the existing BWR design capability.
The principal concerns expressed by the NRC in Reference 2 were also addressed.
The NRC bases for implementation of II.E.4.2(7) as expressed at the meeting were:
(1) the radiation signal would provide additional safety margin, (2) additional redundancy would be provided if one of the existing signals should fail, and (3) there would be additional protection against o n,, !, n a n ? r.
V i sig uk i J N s.
I USNRC Page 2 June 14,1982 low rates of reactor coolant leakage which would not initiate an automatic signal with the current design.
The first two concerns were addressed by the Owners Group presentation on November 19, 1981. That presentation showed that the reliability of the
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existing isolation signals is extremely high.
It also sho. sed that the probability that autgestic isolation on high radiation would be required is extremely low (10 5 for low dose consequence events and 10
- to 10 38 for high dose consequence events).
Such a small benefit would need to be weighed against the high cost of implementation (at least $500,000 per plant for a safety grade system).
With regard to the third NRC concern, the BWR Owners Group has conducted an evaluation for a typical plant of the radiological consequences of the limiting reactor coolant system break which would not result in automatic containment isolation for the current design. The performance of this analysis was discussed with the NRC per Reference 4.
The result is an offsite thyroid dose on the order of.01 Rem which is well below the EPA's Protective Action Guide.
A list of the key input assumptions for that analysis is provided in.
A discussion of the analytical procedure is provided in.
These assumptions and methods will be provided to individual utilities for their use in performing plant unique analyses of the dose-consequences for the limiting break. The NRC Staff has recommended the use of th.e EPA's Protective Action Guide as the acceptance criterion for this event (Reference 3). The BWR Owners Group believes that the dose limit of the EPA's Protective Action Guide is excessively restrictive to be used as a decision basis for installation of an additional automatic isolation signal for. low probability events. However, after reviewing the margin against this criterion for our typical plant analysis, the Owners Group is willing to apply it in this instance to demonstrate that there is no need for automating the high radiation isolation signal.
If a plant unique analysis should show that offsite doses are in excess of this criterion, that utility may elect to adopt a more restrictive technical specification limit on primary coolant iodine concentration for that plant during venting and purging operations so that the acceptance criterion is satisfied. This would be a suitable alternate approach to i
satisfying the intent of NUREG-0737, Item II.E.4.2. Part 7, in lieu of an l
installation of a_n automatic high radiation isolation signal.
In conclusion, the calculation of acceptable offsite doses for the limiting break as defined herein, using the input assumptions and analysis methods defined in Attachments I and 2, is an acceptable alternative to installation of an automatic high radiation isolation signal. Results of r
l a calculation for a typical plant show that this alternate approach is feasible.
Enclosed are copies of relevant correspondence on this subject.
The submittal of an Owners' Group position developed in response to an I
NRC requirement does not indicate that the Owners' Group unanimously endorses that position; rather, it indicates that a substantial number of i
I c
', t.
i-Pag? 3 June 14. 1982 members believe the position is responsive to the NRC requirement and adequately satisfies the requirement.
Each member must formally endorse a position so developed and submitted in order for the position to become the member's position.
Very truly yours, T. J. Dente. 6hairman BWR Owners' Group TJD:rm/A04303 Enclosure cc: BWR Owners' Group J. F. Schilder (GE)
- 5. J. Stark (GE)
D. M. Verre111 (NRC)
W. R. Butler (NRC)
W. Pasedag (NRC)
V. Stello (NRC)
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a.
ATTACMENT 1 Key Analysis Assumptions 1.
Drywell pressure equal to containment isolation setpoint.
2.
Drywell atmosphere is saturated steam.
3.
No plateout or fallout of iodine in containment or vent piping.
4.
No steam condensation in purge or vent pipes.
5.
Break fluid is saturated water at 1000 psia (constant throughout the event).
6.
All iodine in flashed coolant assumed released.
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7.
No credit for standby gas treatment system (SGTS) or reactor water t*
cleanup system (RWCS) filtration.
B.
Initial primary coolant iodine concentration at tech spec limit.
9.
Iodine spiking included (95% cumulative probability value) for depressurization event.
10.
Operator action time to close purge and vent valves = 10 minutes.
s 11.
Annua 1' average, meteorology.
l 12.
Regulatory Guide 1.3 breathing rates.
Conservative assumption, since maximum leaks are on the order of 500 gpm and would be detectable in minutes.
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i FRH:rm/A04304-1 4/30/82
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=
T
- L. : r- :T-~
ATTACHMENT 2 Steam Leakage Rate Calculation The steam leakage rate through the purge or vent line where the pressure drop is small relative to the inlet pressure is:
p)([m.ig4,*]Be,-en-(2'*=T, P,
.w g E k'* l',(M-%3 where:
volumetric flow rate at v, cfs.
=
q g
internal diameter of line, inch.
d
=
pressure, psia.
=
p specific volume of steam, fts/lb,.
=
v, elevation above reference plane, ft.
2
=
resistance coefficient of the line with respect to diameter, K
=
y d.
y (vg3 + vg2)/2 Y
=
g and the subscripts "1" and "2" refer to the conditions at the inlet and outlet of the line respectively.
I I
l l
FRH:rs/A04304-2 4/30/82 I
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p s.,
Break Flow Rate Calculation The flow rate of high pressure saturated water required to produce the quantity of steam, q is F.%s v43 b
(2)
-vX3 where (5' b 4i (3)
K ar IS and break flow rate at vf3, p.
Q
=
b quality, flashing fraction.
x
=
8 specific volume of liquid, ft /lb,.
v
=
f specific enthalpy of liquid, Btu /lb,.
h
=
f h,=
specific enthalpy of vaporization Btu /lb,.
f and subscript "3" refers to conditions in the primary coolant line.
i i
FRH:re/A04304-3 l
4/30/82
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l.,-
Calculation of Activity of Leaking Steam The activity of the steam leaving the vent or purge line at a given time which was produced by the flashing of the high pressure saturated water with activity, a is I
cLf (t-6d p
^*
m.r4 V T-(4) 3 where:
activity leaving the line, Ci/hr.
A
=
activity of high pressure saturated water, pCi/ gram.
a
=
activity of flashed steam per gram partitioning factor =
activity of high pressure saturated f
=
P liquid per gram.
filter efficiency.
c
=
f FRH:rs/A04304-4 4/30/82 a;.
Dose Calculations The offsite dose calculation is as follows:
bi AD;
- Mi 4 i where:
Thyroid dose increment during time interval 1, rem.
=
j Activity released to environment during time interval 1, AA
=
9 Curies (Dose equivalent I 131).
(X/Q), = ' Atmospheric dispersion appropriate to time interval 1 8
Ci-sec/m -Ci.
8 Breathing, rate appropriate to time interval 1, m /sec.
8
=
9 3.47 x 10 4 m /sec for time = 0-8 hr.
8
=
m /sec_for time = 8-24 hr.
8 1.75 x 10 4 (average 2.32 x 10 4 a /see for 0-24 hr.
s Thyroid dose conversion factor for 1131 K
=
1.49 x los rem per Ci-inhaled.
=
l l
l l
I is E
FRH:rm/A04304-5 I
4/30/82 a
.