ML19210B172
| ML19210B172 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 06/18/1975 |
| From: | Lear G Office of Nuclear Reactor Regulation |
| To: | Arnold R METROPOLITAN EDISON CO. |
| Shared Package | |
| ML19210B173 | List: |
| References | |
| NUDOCS 7911040101 | |
| Download: ML19210B172 (21) | |
Text
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NRC PDR Gray file 1
oM j_$
Local PDR j
w o, JUN 1 8 1975 Docket Docket No. 50-289 ORB #5 Rdg GLear KRGoller l
TJCarter OELD Metropolitan Edison Company OISE (3)
ATTN:
Mr. K. C. Arnold DBridges Vice President - Generation SATeets P. O. Box 542 TNovak l
l Reading, Pennsylvania 19603 TBAbernathy l
SVarga Gentlemen:
ACRS (14)
DEisenhut For your preparation of information in response to our December 27, 1974 Order for Modification of Facility License, we have identified generic issues that must be discussed and, where appropriate, resolved by proposed plant modifications and changes to Technical Specifications.
You will find that some of the generic issues identified herein were discussed previously in earlier letters; nevertheless, to assure that your submittal is complete in this respect, we have enclosed a listing and discussion of these additional requirements for information. We have made no attempt l
to adjust this listing and discussion to the unique design and operating features for your f acility; therefore, you must appropriately modify the enclosed docussent for the preparation of your response.
Should yoti have any question ecncerning your implementation of this request for additional information, we will be pleased to advise or to meet with you.
i This request for generic information was approved by CAO under a blanket e.earance number B-180225 (R0072); this clearance expires July 31,,1977.
Sincerely.
/d George Lear, Chief Operating Reactors Branch #3 Division of Reactor Licensing
Enclosure:
Recuest for Additional Information fS 1 - n 7 0 e. :
x cc: w/ enclosure See next page
/
ORB #3
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Form AEC.31s (Rev. 9 53) AECM 0240 W u. s. oovannissut eninvino orricas ser4-sas.ese 2 91104@ fo/ _,
3
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Metropolitan Edison Company JUN 1 8 575 cc:
G. F. Trowbridge, Esquire Shaw, Pittman, Potts. Trowbridge 6 Madden Barr Building 910 17th Street, N. W.
Washington, D. C.
20006 GPU Service Corporation Richard W. Heward, Project Manager Thomas M. Crimmins, Jr., Safety and Licensing Manager 260 Cherry Hill Road Parsippany, New Jersey 07054 Pennsylvania Electric Company Vice President, Technical 1001 Broad Street Johnstown, Pennsylvania 15907 Mr. Weldon B. Arehart, Chairman Board of Supervisors of Londonberry Township 2148 Foxiana Road Middletown, Pennsylvania 17057 Miss Mary V. Southard, Chairman Citizens for a Safe Environment P. O. Box 405 Harrisburg, Pennsylvania 17108 Government Publications taction State Library of Pennsylvania Box 1601 (Education Building)
Harrisburg, Pennsylvania 17126 15C7 074
D REQUIRED INFORMATION 1.
Break Spectrum and Partial Loop Operation,_
The information provided for each plant shall comply with the provisions of the attached memorandum entitled, " Minimum Requirements for ECCS Break Spectrum Submittals."
2.
Potential Boron Precipitation (PWR's Only)
The ECCS system in each plant should be evaluated by the applicant (or licensee) to show that significant changes in chemical concentrations will not occur during the long term after a loss-of-coolant accident (LOCA) and these potential changes have been specif'ically addressed by appropriate operating procedures. Accordingly, the applicant should review the system capabilities and operating procedures to assure that boron precipitation would not compromise long-term core cooling capability following a LOCA. This review should consider all aspects of the specific plant design. including component qualification in the LOCA environment in addition to a detailed review of operating procedures. The applicant should examine the vulnerability of the specific plant design to single failures that would result in any significant boror. precipitation.
3.
Single Failure Analysis A single failure evaluation of the' ECCS should be provided by the applicant (or licensee) for his specific plant design, as r~equired by Appendix K to 10 CFR 50, Section I.D.l.
In performing this evaluatica, the effects of a single failure or operator error that causes any manually controlled, electrically-operated valve to move to a position that could adversely affect the ECCS must be considered. Therefore, if this consid-eration has not been specifically reported in the past, the applicants upcoming submittal must address this consideration.
Include a list of all of the ECCS valves that are currently required by the plant Technical Specifications to have power disconnected, and any proposed plant modifications and changes to the Technical Specifications that might be required in order to protect against any loss of safety function caused by this type of failure. A copy of Branch Technical Position EICSB 18 from the U.S. Nuclear Regulatory Commission's Standard Review Plan is attached to provide you with guidance.
The single failure evaluction should include the potential for passive failures of fluid systems during long term cooling following a LOCA as well as singic failures of active components. For PWR plants, the single failure analysis is to consider the potential boron concentra-problem as an integral part of long term cooling.
4.
Submerged Valves The applicant should review the specific equipment arrangement with-in his plant to determine if any valve motors within containment will become submerged following a LOCA. The review should include all valve motors that may become submerged, not only those in the safety injection system. Valves in other systems may be needed to limit boric acid con-centration in the reactor vessel during long term cooling or may be required for containment isolation.
1567 075
3 The applicant (or licensee) is to provide the following information, for each plant:
(1) Whether or not any valve motors will be submerged following a LOCA in the' plant being reviewed.
(2) If any valve motors will be flooded in their plant, t'he applicant (or licensee) is to:
(a)
Identify the valves that will be submerged.
(b) Evaluate the potential consequences of flooding of the valves for both the short term and long term ECCS functions and containment isolation. The long term should consider the potential problem of excessive concentrations of boric acid in PWR's.
(c) Propose a interim solution while necessary modifications are being designed and implemented.
(currently operating plants only).
(d) Propose design changes to solve the potential flooding problem.
5.
Containment Pressure (PWR's Only)
The containment pressure used to evaluate the performance capability of the ECCS shall be calculated in accordance with the provisions of Branch Technical Position CSB 6-1, which is enclosed.
6.
Low ECCS Reflood Rate (Westinghouse NSSS Only)
Plant., that have a Westinghouse nuclear steam supply shall perform their ECC3 analyses utilizing the proper version of the evaluation model, as defined below:
(1) The December 25, 1974 version of the Westinghouse evaluation model, i.e., the varsion without the modifications described in VCAP-8471 is acceptable for previously analyzed plants for which the peak clad temperature turnaround was identified prior to the reflood Yate decreasing below 1.1 inches per second or for which the reflood rate was identified to remain above 1.0 inch per second; conditions for which the December 25, 1974 and March 15, 1975 versions would be equivalent.
(2) The March 15, 1975 version of the Westinghouse evaluation model is an acceptable model to be used for all previously analyzed plants for which the peak clad temperature turnaround was identi-fied to occur after the reflood rate decreased below 1.1 inches per cecond, and for which steam cooling conditions (reflood rate less than 1 inch per second) exist prior to the time of peak clad temperature turnaround. The March 15, 1975 version will be used for all future plant analyses.
15S7 076
3 APR 25 ms 2
MINIMUM REQUIREMENTS FOR ECCS BREAK SPECTRUM SUBMITTALS I.
INTRODUCTION The following outline shall be used as a guideline in the evaluation of LOCA break spectrum submittals. These guidelines have been formulated for contemporary reactor designs only and must be re-assessed when new reactor concepts are submitted.
f The current ECCS Acceptance Criteria requires that ECCS cooling performance l
be calculated in accordance with an acceptable evaluation model and for a l
number of postulated loss-of-coolant accidents of different sizes, locations l
and other properties sufficient to orovide assurance that the entire spectrum of postulated loss-of-coolant accidents is covered.
In addition, the calculation is to be conducted with at least three values of a discharge i
(C ) applied to the postulated break area, these values spanning coefficient D
the range from 0.6 to 1.0.
Sections IIA and IIIA define the acceptable break spectrum for most operating plants which have received Safety Orders. Sections IIB and IIIB define the break spectrum requirements for most CP and OL case work (exceptions noted later). Sections IIC and IIIC provide an outline of the minimum requirements for an acceptable complete break spectrum. Such a complete break spectrum could be appropriately referenced by some plants. Sections IIID and IIIE provide the exceptions to certain plant types noted above.
l A plant due to reload a portion of its core will have previously submitted all of a break spectrum analysis (either by reference or by specific or part calculations). If it is the intention of.the Licensee to replace expended fuel with new fuel of the same design (no mechanical design differences which could affect thermal and hydraulic performance), and if the Licensee intends to operate the reloaded core in compliance with previously approved Technical Specifications, no additional calculations are required. If the reload core d5 sign has changed, the Licensee shall adopt either of Sections IIA or IIC, or of Sections IIIA or IIIC of this document, as appropriate to the plant type (BWR or PWR). The criterion for establishing whether paragraph A or C shall be satisfied will be determined on the basis of whether the Licensee can demonstrate that the shape of the PCT versus break size curve has not When been modified as a consequence of changes to the reload core design.
the reload is supplied by a source other than the NSSS supplier, the break spectrum analyses specified by Sections IIC or IIIC shall be submitted as a Additional sensitivity minimum (as appropriate to the plant type, BWR or PWR).
studies may be required to assess the sensitivity of fuel changes in such areas as single failures and reactor coolant pump performance.
II.
PRESSURIZED UATER REACTORS Operating Reactor Reanalvses (Plants for which Safety Orders were issued)
A.
If calculational changes
- were made to the LBM** to make it wholly in
- Calculational changes /Model changes--those revisions made to calculational techniques or fixed parameters used tor the referenced complete spectrum.
- LBM--Large Break Modal; SBM--Small Break Model 1587 0,//
R conformance with 10CFR50, Appendix K, the following minimum number of break sizes should he reanalyzed.
Each sensitivity study performed during the development of the ECCS evaluation model shall be individually verified as remaining applicable, or shall be repeated. A plant may reference a break spectrum analysis conducted on another plant if it is the same configuration and core design.
1.
If the largest break size results in the highest PCT:
Reanalyze the limiting break.
a.
b.
Reanalyze two smaller breaks.in the large break region.
2.
If the largest break size does not result in the highest PCT:
Reanalyze the limiting break.
a.
Reanalyze a break larger and a break smaller char. the limiting b.
break. If the limiting break is outside the range of Moody multipliers of 0.6 to 1.0 (i.e., less than 0.6), then the limiting break plus two larger breaks must be analyzed.
If calculational changes have been made to the SBM to make it wholly in small break conformance with 10CFR50, Appendix K, the analysis of the worst (SBM) as previously determined from paragraph C below should be repeated.
i B.
New CP and OL Case Work A complete break spectrum should be provided in accordance with paragraph C below, except for the following:
If a new plant is of the same general design as the plant used as a 1.
basis for a referenced complete spectrum analysis, but operating parameters have changed which would increase PCT or metal-water
~
reaction, or approved calculational changes resulting in more than 20 F change in PCT have been made to the ECCS model used for the referenced the analyses of paragraph A above should be provided complete spectrum, plus a minimum of three small breaks (SBM), one of which is the transition break.* The shape of the break spectrum in the referenced analysis should be justified as remaining applicable, including the sensitivity studies used as a basis for the ECCS evaluation model.
2.
If a new plant (configuration and core design) is applicable to all generic studies because it is the same with respect to the generic plant design and parameters used as a basis for a referenced complete spectrum defined in paragraph C, and no calculational changes resulting in more than 20 F change in PCT were made to the ECCS model used for the referenced complete spectrum, then no new spectrum analyses are required. The new plant may instead reference the applicable analysis.
1587 078
_3 C.
Minimum Requf'ements for a Complete Break Spectrum Since it is expected that applicants will prefer to reference an applicable complete break spectrum previously conducted on another plant, this paragraph defines the minimum number of breaks required for an acceptable complete break spectrum analysis, assuming the cold leg pump discharge is established as the worst break location. The worst single failure and worst-case reactor coolant pump status (running or tripped) shall be established utilizing appropriate sensitivity studies. These studies should show that the worst single failure has been justified as a function of break size. Each sensitivity study. published during the development of the ECCS evaluation model shall be individually justified as remaining applicable, or shall be repeated. Also, a proposal for partial loop operation shall be supported by identifying and analyzing the worst break size and location (i.e., idle loop versus operating loop).
In addition, sufficient justification shall be provided to conclude that the shape of.
the PCT versus Break Size curve would not be significantly altered by the partial loop configuration. Unless this information is provided, plant Technical Specifications shall not permit operation with one or more idle reactor coolant pumps.
It must be demonstrated m.4t the containment design used for the break spectrum anal'ysis is appropriate for the specific plant analyzed.
It should be noted that this analysis is to be performed with an apprc ved evaluation model wholly in conformance with the current ECCS Acceptance Criteria.
1.
LBM--Cold Leg-Reactor Coolant Pump Discharge a.
Three guillotine type breaks spanning at least the range of Moody multipliers between 0.6 and 1.0.
b.
One split type break equivalent in size to twice the pipe cross-sectional area.
c.
Two intermediate split type breaks.
d.
The large-break /small-break transition split.
2.
LBM--Cold Leg-Reactor Coolant Pump Suction Analyze the largest break size from part 1 above.
If the analyses in part 1 above should indicate that the worst cold leg break is an intermediate break size, then the largest break in the pump suction should be analyzed with an explanation of why the same trend would not apply.
3.
LBM--Hot Leg Piping Analyze the largest rupture in the hot leg piping.
79 lpe7 au/
/
4_
4.
SBM--Splits Analyze five different small break sizes. One of these braaks must include the transition split break. The CFT line break must be analyzed for B&W plants. This break may also be one of the five small breaks.
III. BOILING WATER REACTORS The generic model developed by General Electric for BWRs proposed that split and guillotine type breaks are equivalent,in determining blowdown phenomena.
The staff concluded this was acceptable and'that the break area may be considered at the vessel nozzle with a zero loss coefficient using a two phase critical flow model. Changes in the break area are equivalent to changes in the Moody multiplier.
The minimum number of breaks required for a complete break spectrum analysis, assuming a suction side recirculation line break is the design basis accident (DBA) and the worst single f ailure has been established utilizing appropriate sensitivity studies, are shown in paragraph C below. Alsa, a proposal for partial loop operation shall be supported by identifying and analyzing the worst In addition, break size and location (i.e., idle loop versus operating loop).
suf ficient justification shall be provided to conclude that the shape of the PCT versus Break Size curve would not be significantly altered by the partial loop configuration. Unless this information is provided, plant Technical l
Specifications shall not permit operation with one or more idle reactor coolant pumps.
BWR2, BWR3, and BWR4 Reanalysis (Plants for which Safety Orders were issued)
A.
If the referenced lead plant analysis is in accordance with Section III, paragraph C below, the following minimum number of break sizes should be reanalyzed.
It is to be noted that the lead plant analysis is to be performed with an approved evaluation model wholly in conformance with the current ECCS Acceptance Criteria. A plant may reference a break spectrum analysis conducted on another plant if it ~is the same confizuration and core design.
~
Each sensitivity study published during the development of the ECCS evaluation model shall be individually justified as remaining applicable, or shall be repeated.
1.
If the largest break results in the highest PCT:
Reanalyze the limiting break with the appropriate referenced a.
single failure, b.
Reanalyze the worst small break with the appropriate referenced single failure.
Reanalyze the transition break with the single failure and model c.
that predicts the highest PCT.
1587 030
,).
2.
If the largest break does not result in the highest PCT:
Reanalyze the limiting break, the largest break, and a smaller btcak.
a.
If calculational changes have been made to the SBM to make it wholly in conformance with 10CFR50, Appendix K, reanalyze the small break (SBM) in accordance with Section IIIC.
B.
New CP and OL Case Work A complete break spectrum should be provided in accordance with Section III, paragraph C below, except for the following:
1.
If a new plant is of the same general design as the plant used as a basis for the lead plant analysis, but operating parameters have changed which would increase PCT or metal-water reaction, or approved calculational changes have been made to the ECCS model resulting in more than 20 F change in PCT, the analyses of Section III, paragraph A 0
above should be provided plus a minimum of three small breaks (SBM),
one of which is the transition break.
The shape of the break spectrum in the lead plant analysis should be justified as remaining applicable, including the sensitivity studies used as a basis for the ECCS evaluation model.
2.
If a new plant (configuration or core design) is applicable to all generic studies because it is the same with respect to the generic plant design and parameters used as a basis for a referenced complete spectrum defined in paragraph C, and no calculational changes resulting in more than 200F change in PCT were made to the ECCS model used for the referenced ccmplete spectrum, then no new spectrum analyses are required.
The new plant may instead reference the applicable analysis.
C.
Minimum Requirements for a Complete Break Spectrum This paragraph defines the minimum number of breaks required for an acceptable complete spectrum analysis. This complete spectrum analysis is required for each of the lead plants of a given class (BWR2, BWR3, BWR4, BWRS, and BWR6). Each sensitivity study published during the development of the ECCS evaluation model shall be individually justified as remaining applicable, or shall be repeated.
1.
Four recirculation line breaks at the worst location (pump suction or discharge), using the LBM, covering the range from the transition break (TB) to the DBA, including CD coef ficients of from 0.6 to 1.0 times the DBA.
2.
Five recirculation line breaks, using the SBM, covering the range from the smallest line break to the TB.
3.
The following break locations assuming the worst single failure:
a.
largest steamline break b.
largest feedwater line break 1537 03i
. )
largest core spray line break c.
recirculation pump discharge or suction break (opposite d.
largest side of worst location)
D.
BWR4 with " Modified" ECCS Same as Section IIIC.
E.
BWR5 Same as Section IIIC.
F.
BWR6 Same as Section IIIC.
IV.
LOCA PARAMETERS OF INTEREST On each plant and for each break analyzed, the following parameters A.
(versus time unless otherwise noted) should be provided on engineering graph paper of a quality to facilitate calculations.
--Peak clad temperature (ruptured and unruptured node)
--Reactor vessel pressure
--Vessel and downcomer water level (PWR only)
--Water level inside the shroud (BWR caly)
--Thermal power
--Containment pressure (PWR only)
For the worst break analyzed, the following additional parameters B.
(versus time unless otherwise noted) should be provided on engineering The worst single graph paper of a quality to facilitate calculations.
failure and worst-case reactor coolant pump status will have been established utilizing appropriate sensitivity studies.
--Flooding rate (PWR only)
--Core flow (inlet and outlet)
--Core inlet enthalpy (BWR only)
--Heat transfer coefficients 1Ie7
^?
--MAPLHGR versus Exposure (BWR only)
IJV/
bUL
--Reactor coolant temperature (PWR only)
--Mass released to containment (PWR only)
--Energy released to containment (PUR only)
_7_
--PCT versus Exposure (BWR only)
--Containment condensing heat transfer coefficient (PWR only)
--Hot spot flow (PWR only)
--Quality (hottest assembly) (PWR only)
--Hot pin internal pressure
--Hot spot pellet average temperature
--Fluid temperature (hottest assembly) (PWR,only)
A tabulation of peak clad temperature and metal-water reaction (local C.
and core-wide) shall be provided across the break spectrum.
Safety Analysis Reports (SARs) filed with the NRC shall identify on D.
each plot the run date, version number, and version date of the computer model utilized for the LOCA analysis. Should differences exist in version number or version date from the most current code listings made available to the NRC staf f, then each modification shall be identified with an assessment of impact upon PCT and metal-water reaction (local and core-wide).
A tabulation of times at which significant events occur shall be E.
The following provided on each plant and for each break analyzed.
events shall be included as a minimum:
--End-of-bypass (PWR only)
--Beginning of core recovery (PWR only)
--lime of rupture
-Jet pumps uncovered (BWR only)
--MCPR (BWR only)
--Time of rated spray (BWR only)
--Can quench (BWR only)
--End-of-blowdown
--Plane of interest uncovery (BWR only) rn7 pn7
)30/
UO)
Possible grouping of plants for the purpose of performing generic as well as individual plant break spectrum analyses, h
CURRENT DOCKETED APPLICATIONS ~
1537 004
BABC0CK AND WILCOX CATEGORY I:
177 FA w/ Lowered Loops Arrangement Re-analysis (Safety Order Plants)_:
A Ocp5(( '
These plants must resubmit at least 3 breaks.
(They will do Three Mile Island 1 -- IIA so by reference to a complete f
break spectrum reanalysis sub-2535 Arkansas Power 1
-- IIA mitted generically by B&W.)
I 2563 Rancho Seco
-- IIA.
2772 New Ots:
Three Mile Island 2 --IIB (2) i Since these plants are the same 2772 design as the above plant, they Crystal River 3
--IIB (2) (
may reference the same reanalysis 2452 f the complete spectrum above.
Midland 1, 2
--IIB (2) J t
New cps:
None CATEGORY II:
177 FA w/ Raised Loop Arrangement New Ols:
Davis Besse 1
--IIB Complete spectrum required.
New cps Davis Besse 2, 3
--IIB Complete spectrum required.
CATEGORY III: 205-FA Plants New OLs:
None 1537 005
New cps:
IIB i Bellefonte 1, 2 complete spectrum required.
(Plans are for all to reference GrEanwood 2. 3 IIB
(
a complete spectrum submitted IIB f
probablyonWPPSS.)
WPPSS 1, 4 Pebble Springs 1, 2 IIB )
CATEGORY IV:
145-FA Plants New Ols:
None New cps:
IIB Complete spectrum required.
North Anna 3, 4 (One will probably reference IIB J the other.)
Surry 3, 4 O
e 15v7 006
s GENERAL ELECTRIC BWR-2 Oyster Creek
-- LP*
Complete spectrum required.
(IIIA)**
Nine Mile Point
-- Reference only required.
(IIIA)
BWR-3 Quad Cities 2
-- LP*
Complete spectrum required.
(IIIA)**
2511
-- IIIA - 3 _reaks required b
Millstone 2011 Monticello
-- IIIA - 3 breaks required 1670 Dresden 2, 3
-- IIIA 3
2527 h
May reference LP Quad Cities 1
-- IIIA J
2511 Pilgrim
-- IIIA - 3 breaks required 1998 BWR-4 Without fix Hatch 1
-- LP*
Comolete soectrum reauired.
(IIIA)**
2436 Pe Bottom 2, 3 -- IIIA Complete spectrum required. One may reference the other.
Browns Ferry 1, 2, 3 -- IIIA 3293 Cooper
-- IIIA 2381 Fitzpatrick
-- IIIA
['
f3 breaks required.
j Duane Arnold
-- IIIA - 3 breaks required as a reference 1658 Hatch 2
-- IIIA
}
(fortheothers.
2436 g
Brunswick 1
-- IIIA 2436 Shoreham
-- IIIB Fermi
-- IIIB Newbold
-- IIIB
- Lead Plant
- Original break spectrum not wholly in conformance with 10CFR50, Appendix K.
1587 087
~
~
- Comolete spectrum BWR-4 With fi Irunswick 2 (Lead Plant) -- II 2436 required.**
~
Vermont Yankee -- IIIA - 3 breaks required (Lead Plant can be 1593 referenced, if Browns Ferry
- 1, 2, & 3 h appropriate)
Peach Bottom
- 2, 3
)
See preceding page Fitzpatrick*
J BWR-5 Lead Plant
-- IIIE - Complete spectrum required.
Nine Mile Point 2 -- IIIB Complete spectrum required.
i LaSalle 1, 2
-- IIIB (Lead Plant can be referenced by other BWR-5 plants, if Bailly
-- IIIB appropriate.)
Zimmer
-- IIIB Susquehanna 1, 2 --IIIB)
BWR-6 Lead Plant
-- IIIF - Complete spectrum required.
Grand Gulf
-- IIIB Black Fox
-- IIIB 4
Barton 1, 2, 3, 4 -- III.B Complete spectrum required.
(Lead Perry 1, 2
-- IIIB Plant can be referenced by other BWR-6 plants, if appropriate.)
Clinton 1, 2
-- IIIB Douglas Point
-- IIIB Hanford 2
-- IIIB Skagit 1, 2
-- IIIB Hartsville
-- IIIB Somerset
-- IIIB River Bend Station -- IIIB Allens Creek
-- IIIB
- May or may not have the LPCI fix
- Original break spectrum not wholly in conformance with 10CFR50, Appendix K.
000 1r lJu[
UUd m
PLANT SPECIFIC Oyster Creek
-- IIIA Complete spectrum required.
Nine Mile Point
-- IIIA Limerick 1. 2
-- IIIB Hope Creek
-- IIIB Humboldt Bay
-- IIIA Dresden 1
-- IIIA Big Rock
-- IIIA
COMBUSTION ENGINEERING The following list is grouped according to similarities in design.
Some of the older, operating plants are fairly unique, as indicated, and don't fall conveniently into any other groups. The list is in approx, chronological order.
1.
Palisades (Unique) -- IIA 3'
2.
Ft. Calhoun (Unique) -- IIA
)
3 breaks required
- 3.
Maine Yankee (Unique) -- IIA J
4.
2560 MWt Series 3 breaks required a.
Calvert Cliffs Units 1 & 2
-- IIA
)>
Complete spectrum required.
b.
Millstone Unit 2
-- IIB (One may reference the other.)
c.
St. Lucie 1
-- IIB s
- d.
St. Lucie 2
-- IIB Complete spectrum required 5.
3400 MWt Series
( 3410 MWt 217 fuel Assemblies)
- a. Pilgrim 2 (3470Mwt) --IIB \\
Complete spectrum required.
l b.
Forked River 1
-- IIB f
(One may reference the other.)
c.
San Onofre 2 & 3 -- IIB d.
Waterford 3
-- IIB j
6.
Arkansas Class
( 2900 MWt 177 Fuel Assembifes)
~
a.
Russelville 1
-- IIB Complete spectrum required.
(One may reference the other.)
b.
Blue Hills 1
-- IIB s
Maine Yankee is unique in that it has 3 steam generators, 3 hot legs and 3 cold legs. All other CE plants have 2 steam generators, 2 hot legs and 4 cold legs.
- All plants shown above listed before St. Lucie 2 are of the 14x14 fuel design.
All plants after, and including, St. Lucie 2 are 16x16.
r ' : 'l 090 su
IIB Complete spectrum required 7.
System 80 Class-(CESSAR)
These plants have not all b.een named yet. The utility and approx.
number of plants expected are as follows:
a.
Duke (6)
)
b.
WUPPS (1)
May reference complete Arizona Power and Light (2) c.
l d.
TVA (2)
)
l O
i 0
1567 091
l Westinghouse Operating Reactors (Safety Order plants)*
2-loop 3-1000 4-1000 Ginna Surry 1/2 Yankee Rowe Kewaunee Turkey Pt. 3/4 IP2 Pt. Beach 1/2 H. B. Ro'oinson 2 D. C. Cook l Prairie Island 1/2 Zion 1/2 i
Operatina License **
2-1000 3-1000 4-1000 Beaver Valley 1
- Trojan
- Farley 1/2
- Salem 1/2*
North Anna 1/2
- Diablo Canyon 1/2*
IP-3 D. C. Cook 2 McGuire 1/2 Sequoyah 1/2
- 3 breaks required (IIA). One plant may reference another if applicable.
- Complete spectrum required.
One plant may reference another if applicable (see paragraph IIB).
1567 092
Construction Permit **
2-1000 3-loop 4-loop North Coast.
Sharon Harris 1/4 Byron /Braidwood 1/2 Koshkonong 1/2 Catada 1/2 Summer 1 Floating Nuclear 1/8 Beaver Valley 2 Jamesport 1/2 Wisconsin Utilities Seabrook 1/2 SNUPPS 1-5 South Texas 1/2 Comanene Peak 1/2 Watts Bar 1/2 Millstone 3 Vogtle 1/2
- Complete spectrum required. One plant may reference another if applicable (see paragraph IIB).
1507 093