ML20053A500
| ML20053A500 | |
| Person / Time | |
|---|---|
| Site: | Zion File:ZionSolutions icon.png |
| Issue date: | 05/17/1982 |
| From: | Swartz E COMMONWEALTH EDISON CO. |
| To: | Eisenhut D Office of Nuclear Reactor Regulation |
| References | |
| RTR-NUREG-0737, RTR-NUREG-737, TASK-2.B.1, TASK-TM 4129N, NUDOCS 8205260154 | |
| Download: ML20053A500 (10) | |
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ATTACHMENT COMMONWEALTH EDISON COMPAN_Y Zion Station Units 1 and 2 NUREG-0737 Item II.B.1 - Reactor Coolant System Vent - Additional Information 4129N
9.
l.
D Commonwealth Edison
) one First NLtional Plua, Chictgo. filanois
(
C-C' Addrzss Arply to: Post Ottice Box 767 Chicago, Illinois 60690 May 17, 1982 l
l Mr. Darrell
- u. Eisenhut, Director l
Division of Licensing U.S. Nuclear Regulatory Commission Washington, DC 20555
Subject:
Zion Station Units 1 and 2 NUREG 073 7 Item II.B.1 Additional In formation NRC Docke t Nos. $0-295/304 Reference (a):
S.
A. Varga letter to L. DelGeorge dated March 16, 1982.
Dear Mr. Eisenhut:
Reference (a) requested that the Commonwealth Edison Company provide, within sixty (60) days, additional information concerning the reactor coolant system vents at Zion Station.
t attachment to this letter is provided in response to this request.
The To the best of my knowledge and belief, contained in the attachment are true and correct. the statements information furnished by other Commonwealth Edison em In some respects consultants.
Company practice and I believe it to be reliable.Such information has bee may have concerning this matter to this office.Please address any furth One (1) signed original and thirty-nine letter with attachment are provided for your use.(39) copies o f this Very truly yours, y
y -_-
E. Douglas Swartz Nuclear Licensing Administrator At tachtnen t cc:
J. G. Keppler, Region III Region III inspector - Zion lkD 4129N cp \\
8205260l$llb i
T RESPONSES TO NRC REOUEST FOR ADDITIONAL INFORMATION, DATED P14RCH 16, 1982 QUESTION 1:
Your submittal of July 1, 1981, contained information concerning only the reactor vessel head vent.
Provide infornation on the means for venting the pressurizer (reference NUREG-0737 Item ll.B.1 Clarific ation C(3)).
If the existing power-operated (PO"RV) system is designated as the required RCS relief valve for the pressurizer, verify that positive position indi-vent cation for the block valve is provided in the control room (reference NUREG-0737 Item 1: B.1 Clarification A. (5)).
Other -
wise, provide a description of the pressurizer vent and its instrumentation, power and control system, appropriate drawings and'a discussion (with a legend of symbols and abbreviations),
of the pressurizer vent design with respect to each point of clarification of NUREG-0737 Item ll.B.l.
RESPONSE
The existing power-operated relief valve system is designated as the required reactor coolant system vent for the pressurizer.
MOV-RC-8000 Positive position indication for the block valves 1(2) the controls A&B is provided by red and green indicating lights at on the main control board.
QUESTION 2:
Verify that the RCS vent piping inside diameter is smaller than the size corresponding to the definition of a loss-of-coolant accident (10 CFR Part 50, Appendix A) by providing the pertinent design parameters of the reactor coolant makeup system and afrom calculation of the maximum rate of loss of reactor coolant the largest RCS vent break that can be postulated (reference NUREG-07 37 Item ll.B.l. Clarifi' cation A. (4) ).
RESPONSE
The reactor vessel head vent nozzle originally furnished with the reactor vessel and utilized by the reactor vessel head vent system is the restricting line size when considering a loss-of-The internal diameter of coolant accident for the vent systen.At the normal operating conditi'ons this line is 0.742 inches.(Zion FSAR Table 4.1-1) the flow through of 2235 psig and 594.2 F line would be 58.6 lb/sec requiring a makeup flow of 423 gpm this at 120*F.
As 423 gpm exceeds the capacity of charging pumps in the energency core cooling mode, a break of this line must be considered a loss-
.In the event of such a line break, the of-coolant accident.
reactor can be safely ' shutdown through the same means as utilized for any other loss-of-coolant accident as is dicussed in the Tion FSAR, Section 14.3.
~
It should be noted that this 3/4 inch portion of the vent rystem installed as original equipment at Zion.
The new portions was of the vent system have been sized at 1/2 inch which is smaller than the definition of a loss-of-coolant accident in accordance with Section 14.3 of the Zion FSAR.
No new piping was installed for the pressur zer vent system as 7
the existing power-operated relief valves are utilized for this l
system.
QUESTION 3:
In addition to the Westinghouse Ouners' Group generic guidel,ines for operaton of !.he reactor vessel head vent referenced in yo'ur response to NUREC-0737 Item 11.B.1, provide the following additional information:
Procedural guidelines, similar to the Westinghouse 0wners' a.
Group reactor vessel head vent guidelines, for venting of the pressurizer (references NUREC-0737 Item 11.B.1 Position (2) and Cla rification A. (2)).
s
RESPONSE
venting of the pressurrizer will be accomplished via the existing power operated relief valves (PORV's) on the pressurizer.
i Procedures for venting of the pressurizer will be developed on a schedule consistent with the development of the new emergency procedures to be written using the Westinghouse Owners' Group (o f nhich CECO is a member) generic guidelines as a reference.
Th e,
l procedural guidelines for venting of the pressurizer are envisioned to take into account the following factors:
1)
Whether or not Pressurizer Relief Tank.integrityjis intact
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(i.e. whether or not the PRT rupture discs are intact or j
ruptured).
4 2)
Initial hydrogen concentration in the containment priior to l
venting, and the maximum allowable concentration.(p%) to 44 prevent a potentially explosive condition.
3)
Pressurizer level.
4)
RCS pressure.
5)
RCS subcooling.
6)
Operation of Reactor Containment for Coolers to provide i
containment air mixing during venting.
i 7)
Operation of hydrogen control equipment to remove hydrogen from the containment.
l 1
I
4
~e b.
Procedural guidelines which in lieu of venting will assure that suf ficient liquid or steam will flow through the steam generator U, tube region so that decay heat can be e f rectively removed from the reactor coolant system (refarence (NUREG-0737 Item 11.8.1 Cl ari fica tion C. (2) ).
RESPONSE
In response to NUREG 0737 Item 11.B.1 (Clarification C. (2)), no additional venting capability is required other than the vents on the pressurizer and the reactor vessel head.
Zion's present procedure for verification of Natural Circulation (or the Westing-house Owners' Group generic guidelines for verification o f Natural Circulation) is suf ficient to verify that decay heat is being ef fectively removed from RCS.
QUESTION 4:
The following items apply to new portions of the reactor vessel head vent and pressurizer vent that form a part of the reactor l
coolant pressure boundary, up to and including the second normally (reference NUREG-0737 Item ll.B.1 Clarification A. (7)):
closed valve Provide the design tenperature and pressure of any new piping, vc.lves components and supports and verify they are classified a.
Seismic Category 1.
RESPONSE
)
The design temperature and pressure of the new piping and l
valves up to and including the second normally closed valve for the reactor vessel head vent system is 650 *F and 2485 psig (Zion FSAR, Table 4.1-3) ~, respectively.
The new piping, valves, components and supports are classified as Seismic Category 1, s a fe ty-rela ted.
components or supports were required No new piping, valves, for the pressurizer vent system since the existing power-operated relief valve system is utilized.
b.
Verify that any new piping, valves, and components are clas-i sified at least Safety Class 2 (Safety Class 1 where the size corresponds to the 10 CFR Part 50 Appendix A definition of a loss-of-coolant accident.
RESPONSE
All new piping, valves and components installed on the reactor No vessel head vent system are_ classified as Safety Class 1.
new piping, valves or components were installed on the pres-vent system since the existing, power-operated relief surizer valve system has been utilized.
c.
Describe the instrumentation that has been provided to detect and measure pressurizer vent isolation valve leakage (re ference Appendix A to 10 CFR Part 50, General Design Criterion 30).
RESPONSE
The instrumentation that is installed and used to detect leakage past the pressurize.. vent isolation valves consists of the following:
A 110 ohm nickel resistance thermocouple device (RTD) temp-erature element (l (2) TE-4 6 3) is installed in a stainless steel-well loca ted in th a 6 inch pipe downstream from the pres-surizer isolation valves.
This RTD has a calibrated range from 50* to 400'F.
A' temperature indicator is provided on main control room board 1(2)CB06 to provide visual reference to the temperature in the line.
In addition, an audible alarm and an alarm window drop indicator are provided in the main control room.
The alarm will be initiated when the line temp-erature reaches 130 F and will detect leakage that may occur past the valve.
The alarm will reset when She temp-erature drops below 170 F.
No instrumentation is available to directly measure the actual amcunt of this leakage.
We note that measurement of leakage is
-not specifically required by 10 CFR Part 50, Appendix A, General
/
Design Criterion 30.
However, the technical specifications do
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place a limit on the total leakage from the reactor coolant system.
d.
Verify that the materials of construction will be fabricated and tested in accordance with SRP Section 5.2.3,
" Reactor Coolant Pressure Boundary McLerials."
RESPONSE
The materials of construction have been fabricated and tested in accordance with Sargent & Lundy Piping Design Table E-1.
This is the_same piping design table that was used for Ehe rest of the reactor coolant system and Ube materials of con-struction are " discussed in the Zion FSAR, Section 4.2.
- .e i
i a
Nw f
4 Demonstrate $ hat internal missiles and the dynamic ef fects e.
associated with the postulated rupture of piping will not (i.e.,
prevent the essential operation of the vent system r
at least one vent path remains functional) (reference 4
Appendix A to 10 CFR Part 50, General Design Criterion 4).
RESPONSE
Subsection 5.1.2.6, Missile As stated in the Zion FSAR, Protection, the potential sources of internal missiles are:
All valve stems up to and including the largest size to a.
be used.
b.
All valve bonnets.
All instrument thimbles.
c.
d.
Various types and sizes of nuts and bolts.
Complete control rod drive mechanisms.
e.
f.
Reactor vessel head bolts.
A review of the containment piping design drawings identified f
instrument thimbles or miscellaneous no valve stems, bonnets, nuts and bolts that could be considered as potential internal (RVHVS).
missiles af fecting the Reactor Vessel Head Vent System This drawing review was backed up by a physical field inspection for potential missile sources not readily apparent on the The field inspection verified the results of the drawings.
The primary reason this system is not exposed drawing review.
to these types of internal missiles is the shielding provided l
by the refueling pool liner and wall.
t (CRDM) will Failure of a complete control rod drive mechanism not result in a missile that would affect the RVHVS.
The mode of CRDM failure results in a piston type missile which j
would be propelled vertically upward out of its thermal sleeve The RVHVS piping is connection with the reactor vessel head.
routed such that it will not be af fected by such a vertical I
missile.
The reactor vessel head bolts are torqued to the stress level l
Section III.
At this stress level the allowed by ASME Code, stored strain energy level is too low to permit the bolt or stud to become a missile.
I is routed such that it will not be exposed to the The RVHVS dynamic ef fects of a postulated piping rupture.
No new piping or components were required to be installed for the pressurizer vent system since the existing pressurizer power-operated-relief valves are utilized.
QUESTION 5:
Verify that the ilowing reactor vessel head and pressurizer vent failures bs ve been analyzed and found not to prevent the essential operation of safety-related systems required for safe -
reactor shutdown or mitigation of the consequences of a design basis accident:
Seismic failure of any reactor vessel head vent or pres-a.
surizer vent components that are not designed to withstand the safe shutdown earthquake.
RESPONSE
All the reactor vessel head vent and the pressurizer vent (power-operated relief valves) components are designed to withstand the safe shutdown earthquake.
Postulated missiles generated by failure of reactor vessel b.
head and pressurizer vent components.
RESPONSE
The response to Question 4e.above identified the potential sources of internal missiles.
In regard to the reactor vessel head and pressurizer vent components, the potential missile sources are the valve stems, bonne ts, solenoids and nuts and bolts.
A review of the orientation of these items was performed to determine if failure could prevent the essential operation of safety-related systems required for safe shutdown or to mitigate the consequences of a design j
This review, which was backed up with a basis accident.
physical field inspection, concluded that failure of any of the reactor vessel head or pressurizer vent components will not prevent the essential operation of safety-related i
systems required either for safe chutdown or to mitigat'e the consequences of a design basis accident.
Dynamic ef fects associated with Ohe postulated rupture of c.
new pressurizer vent piping greater than one inch nominal size.
RESPONSE
No new pressurizer vent piping was installed.
r 4
Fluid sprays from reactor vessel head and pressurizer vent J
d.
from normally unpressurized component failures.
Sprays portions of the vents that are Seismic Category 1 and Safety 2 or 3 and have instrumentation for detection of Class 1, from upstream isolation valves need not be considered.
leakage
RESPONSE
A review was conducted to identify any safety-related equip- -
ment required either for safe shutdown or to mitigate the consequences of a design basis accident that could be af fected by fluid spray from reactor vessel head and pressurizer vent The results of this review were verified component failures.
The review concluded that no by a physical field inspection.
safety-related equipment required either for safe shutdown or to mitigate the consequences of a design basis accident would be prevented from operating.
It should be noted that a deflection shield was mounted on the CRDM seismic platform. between the solenoid valves and the control rod drive mechanisms and associated wiring, to pre-vent any water (valve leakage or line breakl from af fecting the drive mechanisms.
QUESTION 6:
Demonstrate that the pressurizer vent paths to the containment atmosphere discharge into areas:
That provide good mixing with containment air to prevent the accumulation or pocketing of high concentrations of hydrogen, a.
and
RESPONSE
The pressurizer vents to the containment atmosphere through the These rupture discs are pressurizer relief tank rupture discs.
located on the top of the pressurizer relief tank adjacent to the reactor containment fan ducts from which approximately 5000 CFM of air is discharged into the area of the rupture disc.
that the accumulation or pocketing I
This flow is adequate to ensure of high concentrations of hydrogen will be prevented, i
systems, and components essential In which any nearby structures, to safe shutdown of the reactor or mitigation of a design basis b.
accident are capable of withstanding the effects of the anticipated -
liquid, and noncondensible gas disch,arging mixtures of steam, (reference NUREG-0737 Item ll B.1 l
from the pressurizer vent j
Clarification A. (9)).
l l
RESPONSE
The pressurizer relief tank rupture disc vents upward from a dis-f charge elevation of 579'-0" to the bottom of the elevation 590 '-
There is no sa fety-related equipment in the 0" concrete slab.
The environmental conditions that will direct discharge path.
from such a discharge are less severe than those that would be present af ter th'e worst case guillotine loss-of-coolant result
.,pe
r. _
accident.
Since the equipment required for safe shutdown of
~
the reactor and to mitigate the consequences of a design basis accident is qualified to withstand the more severe post LOCA environment, they will be capable of withstanding the ef fects of the anticipated mixtures of steam, liquid and gasses discharg-ing from the pressurizer relief tank rupture disc.
QUESTION 7:
Verify that operability testing of the reactor vessel head and pressurizer vent valves will be performed in accordance with Subsection IWV o f Section XI o f the ASME Code for Category B Valves (reference NUREG 0737 Item 1 I.B.1, Clarification A. (ll)).
RESPONSE
Reactor Vessel Head Vent System - When operational, operability testing o f the system will conform to ASME Section XI (1974 to summer 1975 addendum), Subsection IWV-3000 with the following considerations from Subsection IWV-3410(b):
The valves will be stroke tested each cold shutdown provided all reactor coolant pumps are not operating.
The testing period will be each refueling outage as a maximum; and, in case of frequent cold shutdowns these valves need not be exercised more often than once every three months.
Pressurizer Vent Valves - The power operated relief valves and block valves testing conforms to the ASME Section XI Subsection IWV with the exception o f valve stroke timing (IWV-3410(d)).
Zion Station will augment the testing requirements to comply with Section XI.
QUESTION 8:
Verify that all displays (including alarms) and controls, added to the control room as a result of the TMI Action Plan Requirement for reactor coolant system vents, have been or will be considered in the human f actors analysis required by NUREG-0737 Item 1.D.1,
" Control Room Design Reviews," in addition to their consideration in finalizing the station emergency procedures.
PESPONSE:
i Human factors engineering contractor, Station Electrical Engineering, Station Nuclear Engineering Department, Sa rgent &
Lundy and Station personnel were involved in review of and final location for displays, alarms and controls added to control room for RCS vents.
i
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