ML19296D273
| ML19296D273 | |
| Person / Time | |
|---|---|
| Site: | Big Rock Point File:Consumers Energy icon.png |
| Issue date: | 02/22/1980 |
| From: | Gallo J CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.) |
| To: | BIER, MILLS, CHRISTA-MARIA, ET AL |
| Shared Package | |
| ML19296D272 | List: |
| References | |
| NUDOCS 8003030167 | |
| Download: ML19296D273 (70) | |
Text
.f.
9 2/22/80 UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of
)
)
CONSUMERS POWER COMPANY
)
Docket No. 50-155
)
(Big Rock Point Nuclear
)
Power Plant)
)
ANSWERS OF CONSUMERS POWER COMPANY TO INTERROGATORIES PROPOUNDED BY CHRISTA-MARIA, ET AL.
Pursuant to 10 C.F.R.
S 2.740b and the schedule for discovery set forth in the Atomic Safety and Licensing Board's " Order Following Special Prehearing Conference,"
Consumers Power Company
(" Licensee") hereby submits answers to Interrogatories 2-14, 7-1 through 7-11, and tha portion of 2-10 that was deferred (see Licensee's response filed on February 20, 1980).
Interrogatory 7-1 Please explain whether there is or will be any airborne radiation in the atmosphere above the spent fuel pool (but within the containment) as a result of spent fuel pool opera-tion, including fuel movement and pool maintenance activities, and including all credible accidents.
What is the source of the airborne radiation?
What type of radiation is it?
A.
Answer:
Radioactive material exists within spent fuel in particulate, gaseous and halogen form.
In addition smaller amounts of particulate radioactive material exist on the exterior surfaces of the spent fuel.
3003030 hbk
. During normal operation of the spent fuel pool including fuel movement and maintenance activities small alaounts of the gaseous radioactive material from within the fuel (krypton and xenon), halogen (iodine) as well as the particulate radioactive material on the exterior surfaces of the spent fuel can be expected to be released to the pool water.
These radioactive materials within the spent fuel pool water create measurable radiation fields above the spent fuel pool water surface.
At approximately 10 feet above the spent fuel pool water surface the radiation field created by these radioactive materials in the spent fuel pool water is not distinguishable from radiation fields caused by other radioactive sources within the containment.
The radiation levels from the radioactive materials in the spent fuel itself are imperceptible at the pool water surface.
Of the radioactive materials in the spent fuel pool water, only the gases (xenon and krypton) and small amounts of radioactive iodine can potentially become airborne to the atmosphere above the pool.
In additior.
small a.
1.ts of particulate radioactive material existing on internal surfaces of the cor.tainment build-ing near the spent fuel pool may be disturbed and become airborne as radioactive dust as a result of maintenance activities.
. During accidents, including those mentioned on page 4.1 of the Company's Environmental Impact Evalu-ation of the Big Rock Point Plant Spent Fuel Rack Addi-tion, gases and radioactive iodine can be expected to be released to the containment atmosphere if spent fuel cladding is breached.
The source of radiation, whether during normal operation or accident conditions, is radioactive decay of gaseous, halogen and particulate radioactive material within the water of the spent fuel pool and on internal surface of the containment building.
The radiation itself is composed of alpha and beta particles and gamr..a rays.
Table I (column entitled " shutdown") shows the amount of radioactive material in spent fuel which has been in the reactor for three years, in curies per thermal megawatt.
To obtain the radioactivity levels at any particular decay time in a single bundle of Big Rock Point spent fuel, one should multiply the numbers in the table by approximately 2.7 thermal megawatts.
As the table shows, there are no significant quantities of gaseous and iodine radioactive material which might become airborne after one year after discharge.
. The foregoing answer responds to that part of Interroga:.uiy 2-10 concerning radiation over the spent fuel pool given the existing pool configuration.
B.
Documents Relied Upon Oak Ridge National Laboratory computer code called "ORIGEN" which provides an inventory of radioactive material by isotope resulting from uranium fission.
R. W. Sinderman work papers for Table I (copy enclosed)
Environmental Impact Evaluation, dated April 1979 (pages 3-2 and 4-1) for the spent fuel rack addition for the Big Rock Point Plant.
U.S.
NRC Regulatory Guide 1.25 C.
M.
- Lederer, J.
M.
Hollander and I.
Perlman, Table of Isotopes, 6th Ed. (1967)
C.
Documents Reviewed But Not Relied Upon R. W. Sinderman, Statement Before the Special Joint Committee on Nuclear Energy (1/21/80) of the Michigan legislature.
(copy enclosed)
Attachment A (postulated cask drop accident study) to letter of R.
B.
Sewell to Karl R.
Goller, dated July 1, 1974 (a copy of this document will be furnished in response to Interrogatory 9-37).
D.
Further Activities None.
Table I Tission asd Transura ius Radicactive Material in Spent Fuel As a Tu=ctics of recay Time in Curies Fer 'her=al Megawatt
- 100 1000 10,000 EE
..a Shutdeva 1 Ecur 1 Cay 10 Oays 1 Month 1 Year 10 Years Years Years Years
'ine 2
Callium 9T Oe:=aniu=
296 Arsenic h,617 Seleci=
21,730 L65 9
.03
.01
'.01
.01
.01
.01
.01 3rc=ine 61,080 3,5c6 Krypten 101.100 28,270 Rubidius 168,600 1,331,000 5,003 Strontiu=
19h,700 76,590 30,030 22,150 17,' 30 2,559 1,921 202 Yttrius 283,600 12k,500 hh,960 30,020 2k,2ho 2,83h 1.921 253 Zireer.ius 176,600 90,830 63,210 hl,850 33,810 95h
.10
.10
.10
.10 Niobiu 33k,500 138,500 83,080 k7,520 LL,350 2,050 Molybde==
22h,900 53,910 39,270 h,063 Technetiu=
298,900 89,000 37,780 3,93h 26
.5
.5
.5 5
.5 Ruthe:1=
130,200 90,910 59,990
$2,970 L2,*.60 9,6h8 19 Rhodium 168,100 106,200 80,850 52,670 hl,700 9,6L7 19 Palladi=
n,L50 8,523 2,725 Silve.r ik,700 lo,1ho h,225 Cad =iu=
1,k30 Indium 232
-~
~
Tin 36,610 5,123 61h 262 93 7
.05
.03
.01
.01 Anti =c 7 117,100 2h,850 3,kT8 1,101 E92 Tellurium 202,300 99,130 k3,570 8,168 2,103 1h2 Iodine 313,800 181, Loo Sh,h60 12.550 2,222 Xenos 2h3,100 78,690 61,k90 17,020 1,265 Cesium 207.h00 788, LOO 6,369 6,026 5,6h2 h,892 2,90h 355 Barium 235. Loo 112,h00 k8,150 30,860 12,650 3,266 2,653 332 Lanthanu=
216,000 127,300 h9,h90 31,600 10,720 Cerium 196.000 126,600 109,200 Th,650 58,770 1h,h60 Praseody=ius 158,200 113,900 78,710 62,080 L2,710 1k,hko Neodymiu=
32,560 2h,010 15,800 9,007 2.583 Prc=ethiu=
35.580
-)0,030 23,810 8,787 T,096 5.288 h9 Se=arius 9,632 8,270 5,718 Europium 32 Gadclinius 169 Terbius 60 Dysprosium 7
.5 Total Fission h,191,000 3,866,000 9h7,Lc0 520,h00 350,600 70,700 9,610 1,1h5
.6
.6 Neptunius
.008
.008
.008
.008
.008
.008
.008
.008
.008
.008 Plutoniu=
h,689 h,689 h,689 h,689 h,689 h,689 2,950 8k 2h 13 Ame '.cium 6
6 6
6 6
6 6
5 2
3 Curic=
101 101 101 101 101 101 71 2
Total Transuranie k,796 h,796 h,796 h,796 h,?96 h,790 3.030 91 26 13 Total h,196,000 3,871,000 952,200 525,200 355.h00 75.500 12,6ho 1,236 27 1h
' Cal Ridge National laboratory ORICEN Cocputer Code for Decay times to 1 year for fissicn products. All others
. Interrogatory 7-2 How are airborne radiation levels in the atmosphere above the spent fuel pool (but within the containment) monitored?
Please provide a complete description of the monitoring that has been done to date, including complete records of the airborne radiation levels.
A.
Answer:
The airborne radiation levels in the containment are monitored by (1) two continuous air monitors (CAMS) which continuously monitor containment air, with strip chart readout, and which alarm at preset radiation levels; (2) five area monitors which continuously monitor radiation fields within containment, with multipoint recorder strip chart readout in the control room and which annunciate in the control room at preset radiation levels; and (3) grab samples of containment air taken approximately daily for Plant laboratory analysis of airborne radiation levels.
The two continuous air monitors (CAMS) are located inside containment.
During normal operation, one CAM is located near the personnel lock which is about 50 feet south of the south wall of the spent fuel pool and about 16 feet below the level of the surface of the spent fuel pool.
The second CAM is located on the discharge of the exhaust plenum which is at the opposite side of the containment from the spent fuel pool.
This CAM monitors the airborne activity in containment air flow before the flow exits the containment and is released to the environment through stack.
During refueling operations one CAM is placed on the reactor deck level near the area monitor on the steam drum wall to monitor refueling operations.
If fuel sipping operations are required, a CAM is placed about 10 feet north of the spent fuel pool to monitor the fuel sipping operation which occurs in the spent fuel pool at the east end of the pool.
Two area monitors are located near the spent fuel pool.
One is located at the south west corner of the spent fuel pool edge and stands approximately 10 feet above the surface of the pool.
In addition to the con-trol room readout and alarm this monitor has a local readout (dial indication) and a local alarm.
The monitor indicator normally reads 8-9 mrem /hr. and has an alarm setting at 20 mrem /hr.
The second area monitor is located on the steam drum wall approximately 35 feet southwest from the area monitor located at the spent fuel pool edge.
Again, in addition to a control room readout and alarm this area monitor also has a local readout and a local alarm.
It normally reads 12-13 mrem /hr.
This interrogatory also requests complete records of airborne radiation levels monitored within containment to date.
This information has been compiled for the 18 years of reactor operation at Big Rock Point.
It consists
. of strip charts and grab sample analysis data sheets.
The two monitors (CAMS) generate three inches of data per hour on a continuous basis.
Thus:
3 in/hr x 24 hr x 365 day x 18 yrs x 2 monitors =
946,080 inches or approximately 15 miles of paper There is a combined strip chart for the area monitors which also generates approximately 3 inches of data per hour.
This information except for the most recent year is stored in the Licensee's record repository in Traverse City, Michigan.
The most recent information is located at the Big Rock Point Plant.
With respect to the grab sample analysis data sheets, one such sheet is generated per week.
This means (1 sheet x 52 weeks x 18 years) there are approx-imately 936 sheets.
The material has been reduced to microfilm (excep; for the most recent year) and it is stored in Traverse City, Michigan.
The most recent information is located at the Big Rock Point Plant.
Licensee objects to this part of Interrogator;r 7-2 because it is unduly burdensome to provide the large quantity of records indicated above.
In addition to the vast amount of documentation, the strip charts are on a continuous roll and are not readily reproduced.
Upon request, Licensee will make this information available for inspection by Christa-Maria's counsel at the storage location at a mutually convenient time.
-9_
B.
Documents Relied Upon C.P.Co. Drawing No. 740F30762 Rev. C (copy enclosed)
C.
Documents Reviewed But Not Relied Upon None.
D.
Further Activities None.
Interrogatory 7-3 Does Licensee project any increase in the above airborne radiation levels as a result of increased fuel storage?
If so, under what conditions?
If not, please justify your answer.
A.
Answer:
Increases in airborne radiation levels as a result of increased fuel storage are expected to be imper-ceptible.
This is true for both normal operations and accident conditions.
Increased fuel storage capacity will not result in any increase in the amount of frechly off-loaded fuel existing in the spent fuel pool at any particular time.
Thus, the additional capacity would accommodate the spent fuel with the oldest decay time.
Kr-85 is the only radionuclide remaining in spent fuel having decayed for a year or more which is poten-tially releasable to the containment atmosphere.
During normal operation of the spent fuel pool, Kr-85 release will be negligible.
For credible accident conditions described in Interrogatory 9-19 where cladding is breached, Kr-85 releases from decayed spent fuel bundles after expansion of the pool capacity are no greater than releases of Kr-85 from such bundles prior to such expansion.
With respect to that part of Interrogatory 2-10 concerning radiation over the spent fuel pool if the capacity of the pool is expanded, no different radia-tion levels than discussed in the first paragraph of page 2, are expected.
B.
Documents Relied Upon Environmental Impact Evaluation, dated April 1979 (page 4-1) for the spent fuel rack addition for the Big Rock Point Plant.
C.
Documents Reviewed But Not Relied Upon Attachment A (postulated cask drop accident study) to letter of R.
B.
Sewell to Karl R.
Goller, dated July 1, 1974 (copy to be enclosed with answer to Interrogatory 9-37).
D.
Further Activities None.
Interrogatory 7-4 If Licensee projects any increase in the above airborne radiation levels, please explain the projection and provide all technical calculations and other supporting material.
A.
Answer:
Since increases in airborne radiation as a result of increased fuel storage are not perceptible, no calcu-lations have been performed.
B.
Documents Relied Upon None.
C.
Other Documents None.
D.
Further Activities None.
Interrogatory 7-5 How has the above airborne radiation been removed from the con-tainment atmosphere to date?
How does Licensee intend to remove it after the spent fuel pool storage capacity has been expanded?
A.
Answer:
All airborne radioactive material within the con-tainment including any which is generated as a result of spent fuel pool operation is removed from the containment by continuous ventilation to the plant stack.
This ven-tilation system provides a continuous flow across the surface of the spent fuel pool to quickly remove any airborne radiation to the stack.
No changes in this mode of operation are anticipated after the spent fuel storage capacity has been expanded.
B.
Documents Relied Upon Big Rock Point Plant Final Hazards Summary Report, Section 6.8.4, Page 8.
(copy enclosed)
Big Rock Point Plant Piping and Instrument Diagram M-125, Rev. N, and M-124, Rev.
G.
(copies enclosed)
C.
Documents Reviewed But Not Relied Upon None.
D.
Further Activities None.
Interrogatory 7-6 Please explain all of the possible pathways that the above airborno radiation could follow, including any ways that the radiation could be released to the atmosphere outside the containment.
A.
Answer:
Pathways that airborne radioactive materials within containment could follow are:
1.
Deposition of airborne particulates and halogens onto surfaces and their subsequent radioactive decay within the containment.
2.
Release of airborne particulates, halogens and noble gases to the atmosphere through the plant stack via the containment ventilation system.
B.
Documents Relied Upon Big Rock Point Plant Piping and Instrument Diagram M-124, Rev.
G, and M-125, Rev. N.
Big Rock Point Plant Daily Radiation Surveys, 1980.
Big Rock Point Stack Gas Sample Filter Analysis, 1979.
C.
Documents Reviewed But Not Relied Upon None.
D.
Further Activities None.
1
' Interrogatory 7-7 Assuming the spent fuel pool to be filled to capacity, including a recent addition of a full core, what would be the level of airborne radiation in the atmosphere above the spent fuel pool (but within the containment) ?
A.
Answer:
Records indicate that levels of airborne radioactivity near the spent fuel pool before refueling (with decayed spent fuel in the spent fuel pool) are not significantly different from those present after a fresh core off-load (84 bundles).
Records for 1976 and 1978 (a refueling operation did not occur in 1977) indicate that generally airborne radioactivity in the vicinity of the spent fuel pool
-10 average 2.3x10 uCi/cc during normal plant operation
-10 and 3.4x10 uCi/cc with a recent full core off-load.
For the reasons stated in the answers to Inrcrrogatories 7-1 and 7-3, it is expected that these levels will not increase when the spent fuel pool is filled to capacity.
B.
Documents Relied Upon Weekly Radiological Survey Air Sample Data Sheets which were condensed into a one page data sheet (copy enclosed).
C.
Documents Reviewed But Not Relied Upon None.
D.
Further Activities None.
Interrogatory 7-8 Provide all projections of increased dose rates in the atmosphere outside the containment as a result of the in-creased storage of spent fuel.
Justify your answer.
A.
Answer:
No increases in present radiation doses outside the reactor building at the site boundary are expected due to airborne cadioactivity or direct radia: ion as a result of increased amounts of spent fuel stored within the spent fuel pool.
Direct radiation penetrating the pool water, pool concrete walls and containment building is presently below background levels outside the contain-ment building and is not expected to increase as a result of increased storage as more fully explained in the responses to interrogatories relating to Contention 2.
Increase in radiation dose rates as a result of releases of radioactive material to the containment and ultimately the outside atmosphere are not expected to increase, as explained in the response to Interrogatory 7-3.
B.
Documents Relied Upon None.
C.
Documents Reviewed But Not Relied Upon None.
D.
Further Activities None.
. Interrogatory 7-9 What is the role of the containment itself in preventing the escape of airborne radiation from the spent fuel pool into the outside atmosphere?
A.
Answer:
During normal plant operation, the containment itself does not function to prevent the escape of air-borne radioactive material from the spent fuel pool into the outside atmosphere.
The containment provides a controlled pathway for such releases through the plant stack which is continuously monitored by the stack gas sample and monitoring system.
During accident conditions, the role of the contain-ment is to provide an essentially leak-tight boundary to prevent the spread of radioactive material to the outside environs.
For the specific case of an accident causing the release of airborne radiation from the spent fuel pool, an alarm by either of the two contain-ment area monitors in the vicinity of the spent fuel pool described in the response to Interrogatory 7-2 will automatically cause the containment to isolate, thereby preventing the escape of significant quantities of airborne radiation.
The containment will also isolate automatically upon other signals, not necessarily related to spent fuel pool accidents, such as high containment pressure or reactor scram.
The containment may also be isolated by remote operator action from the control room.
The co:.tainment design maximum leakage rate at 27 psig is 0.5% of the containment air volume per day.
The containment is tested for leak-tightness in accordance with 10 C.F.R. Part 50, Appendix J.
B.
Documents Relied Upon Final Hazards Summary Report (FHSR), Section 3.2.2 (copy enclosed).
C.
Other Documents Reviewed But Not Relied Uoon C.P.Co. Drawing 0740G40124 Rev.
G.
C.P.Co. Drawing 0740G40125 Rev.
N.
C.P.Co. Big Rock Point Plant Emergency Procedure EMP 3.4, Rev. 3 " Fuel Damage while Refueling" (copy enclosed).
FHSR, Section 3.1 (copy enclosed).
FHSR, Section 6.8.4 (copy enclosed).
United States Nuclear Regulatory Commission Office of Nuclear Reactor Regulation Report on the Systematic Evaluation of Operating Facilities (dov. 25, 1977) (copy furnished to Christa-Maria by NRC Staff).
Letter From Richard H. Vollmer, USNRC to David P.
- Hoffman, C.P.Co.,
Re Status and Categorization of Systematic Evaluation Program (SEP) Topics (Dec. 21, 1979) (copy enclosed)
D.
Further Activities The NRC Staff, in its systematic evaluation program, is conducting a generic review of, among other things, the following four topics:
- SEP Topic III-7.D Containment Structural Integrity Test SEP Topic VI-2 Containment Functional Design SEP Topic VI-4 Containment Isolation System SEP Topic XV-20 Radiological Consequences of Fuel Damaging Accident (Inside and Outside Containment)
The Licensee has not yet undertaken any activities relating to these topics for the Big Rock Point Plant.
Interrogatory 7-10 Have any questions been raised concerning the ability of this containment or similar containments to per'crm that function?
Please explain.
A.
Answer:
Although a detailed searcu of the Big Rock Point Plant Docket and over 20 years of J21ated corraspondence has not been mace, the persons responding to this luserrogatory believe that if such questions had ever been communicated to Licensee or raised by Licensee, it is likely that they would be aware of them.
To the best of their knowledge, information and belief, no verbal or written questions concerning the adequacy of the Big Rock Point Plant containment or similar contain-ments to perform the functions described in the response to Interrogatory 7-9 have ever been communicated to Licensee or raised by Licensee.
However, the NRC Staff is conducting further activ-ities in connection with the Systematic Evaluation Pro-gram as discussed in Part D of our response to Inter-rogatory 7-9.
. Further, as a result of the Three Mile Island accident, the NRC Staff has requested a review of the Big Rock Point Plant shielding design requirements as described in NUREG-0578, Section 2.1.6.b.
The Licensing Board in this preceeding has asked a similar question.
See Consumers Power Company (Big Rock Point Nuclear Plant), Order Following Special Prehearing Conference, Docket No. 50-155 (January 18, 1980), at p.
28.
How-ever, these questions concerning the adequacy of the containment to provide biological shielding from direct radiation are different from the questions in Inter-rogatories 7-9 and 7-10 concerning the adequacy of the containment to prevent escape of airborne radiation from the spent fuel pool.
B.
Documents Relied Upon NUREG-0578, Section 2.1.6.b C.
Documents Examined But Not Relied Upon See documents listed in Section C of the response to Interrogatory 7-9.
D.
Further Activities See Section D of the response to Interrogatory 7-9.
. Interrogatory 7-11 In a letter dated October 19, 1979, the Licensee responded to staff questions issued September 4, 1979.
Question #2 concerned leak collection and detection from the spent fuel pool.
The Licenree explained that "[p]eriodic inspections" were made at collection lines which terminated at an "open basin" and drained into the reactor building sump and the liquid radwaste system.
How often are such " periodic inspections" made and by a.
whom?
b.
Does the open collection basin hold any leakage that may occur, or does this basin simply drain into the reactor building sump?
Is it possible to observe directly the liquid level of c.
the open collection basin?
d.
The Licensee stated that "[m]oisture observed during periodic inspection is believed to be due to condensa-tion rather than pool leakage."
What is the basis of this belief?
Is there a chemical distinction between such condensation and spent fuel pool water?
Have there been any evaporation losses from the open e.
basin?
If so, what has happened to the resulting air-borne radiation?
A.
Answer:
Periodic inspections are made by Big Rock Point Plant a.
operations personnel at 90 day intervals in accordance with a written procedure, T90-68, entitled " Spent Fuel Pool Liner Leak Rate Test".
b.
The open basin is simply a sink below the point where eight collection lines terminate.
Each collection line ends in a valve which is normally closed.
During the leakage test, these valves are opened one at a time and any moisture collected from a collection line is recorded.
The sink is not designed to hold water and drains into the reactor building sump.
. c.
Any moisture in the sink would be obser able, but as explained above the basin is not designed to hold water.
d.
The spent fuel pool is a double walled container.
The inner wall is stainless steel and the outer wall is coated concrete.
At the floor of the pool, between the two walls, are eight cavities which have drainage collection lines that are routed to the open basin described above.
The spent fuel pool was originally built without a stainless steel liner.
After the liner installation in 1974, leakage tests similar to those described above were conducted on a weekly basis.
After sufficient tests had been conducted to confirm that only negligible amounts of water were being collected, the tests were rescheduled on a 90 day basis.
Moisture has never consistently shown up from the same collectior. line, and therefore it has been assumed that the drops observed from time to time are attributable to condensation that occurs in the space between the liner and the concrete walls.
There is a chemical distinction between condensation of airborne maisture and spent fuel pool water, which is chemically very pure.
However, it is unlikely that analyzing the moisture from the collection lines would identify whether the origin of the moisture was leakage
- or condensation.
This is because any water will pick up impurities from the concrete walls and floor of the pool as it trickles slowly to the collection lines.
It might be possible to distinguish between leakage and condensation in that the leakage would have a larger proportion of short-lived radioactive isotopes.
e.
No.
See t'a answers to b and c above.
B.
Documents Relied Upon Big Rock Point Plant procedure T90-08, Rev.
1,
" Spent Fuel Pool. cer Rate Test" (copy enclosed).
C.
Documents Examined Relied Upon None.
D.
Further Activities C. Axtell intends in the near future to con. pare the isotopic contents of moisture obtained from the collec-tian lines with spent fuel pool water for the comparison described in the last sentence of "d" above.
Interrogatory 2-14 Explain in detail why le Licensee anticipates that expansion cf the spent fuel pool capacity by over 100 percent will cause "no increase in dose rates over..rse previously experienced due to radio-nuclides.
at the edge of the pool."
(Design and safety analysis, P.
7-1).
What are those dose rates and precisely where and how are they measured?
A.
Answer:
The storage of additional fuel in the pool will not cause an increase in the equilibrium radionuclide con-centrations in the pool water.
The spent fuel pool cleanup system has the ability to maintain radionuclide
- concentrations in the pool which are independent of the amount of fuel stored.
Therefore, no increase in dose rates over those previously experienced due to radio-nuclides is expected at the edge of the pool.
See also responses to Interrogatories 7-1, 7-2, and 7-3 and 2-15.
The dose rates in the vicinity of the pool as measured with a portable radiation monitor are given in two survey maps attached to a handwritten memorandum from C.
E.
Axtell to C.
L.
Larsen, dated February 5, 1980.
P.
Documents Relied Upon Description and Safety Analysis Report, dated April 23, 1979, p.
7-1.
Table 1, attached to the answer to Interrogatory 7-1.
- Memo, C.
E.
Axtell to C.
L.
Larsen, 2/5/80.
Documents listed in answers to Interrogatories 7-1, 7-2, and 7-3.
C.
Documents Reviewed But Not Relied Upon None.
D.
Further Activities None.
b Jo ph gallo' of the Attorneys for Consumers Power Company Isham, Lincoln & Beale 1050 17th Street, N.W.
Suite 701 Washington, D.C.
20036 202/833-9730
District of Columbia) SS.
UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFET'.' AND LICENSING BOARD In the
'n ar of
)
)
CONSUMERS POWER COMPANY
)
Docket No. 50-155
)
(Big Rock Point Nuclear Power Plant))
AFFIDAVIT OF CARL L.
LARSEN I,
Carl L.
Larsen, of lawful age, being first duly sworn, do state as follows:
I am employed by Consumers Power Company as an engineering supervisor in the Generating Plant Modifica-tions Depa.' tment.
I have overall responsibility within the Company for technical, cost, and schedule aspects of the proposed spent fuel pool expansion at the Big Rock Point Plant.
My resume is attached.
I have primary responsibility for answering Interrogatory 7-9.
I have joint responsibility for the response to Interrogatory 7-10.
To the best of my knowledge and belief, the statements in this affidavit and in the responses to the interrogatories listed above are true and correct.
7 C4LN i
A
~
Car 3/I/.l Larsen j
-l Subscribed and sworn to before me this 22nd day of February, 1980 I til:L -
l b.*
Notary Public //
My Ccmminien Eq > cepember 14, 1982
s.
CARL LEE L1RSEN EXPERIENCE:
Consumers Power Co=pany since 1979 as Project Manager, responsible 1979 to for several major modifications to operating nuclear power plants.
Present Responsibilities include technical, cost and schedule aspects in-clu 'ang vendor selection and construction interface.
1974 to 1979 Gilbert /Cocoonwealth Associates, Inc.1973 to 1979.
1978-1979 Senior Licensing Engineer responsible for the preparation of security system design descriptions for a nuclear power plant, responses to USNRC questions on fire protection systems and technical support for hearings before the ACRS concerning a CP stage license application for a nuclear power plant.
1976 - 1978 Responsible for the preparation of responses to USNRC questions during CP review of PSAR license application.
Participated in technical meetings with USNRC Staff regarding USNRC questions on PSAR.
1975 - 1976 Lead Safety Licensing Engineer responsible for coordinating the preparation of a Preliminary Safety Analysis Report for a nuclear power plant.
Responsibility for the licensability of technical information for the PSAR.
Coordinated the preparation of plant security docu=ents and fire protection and emergency planning.
Performed NSSS vendor evaluation and liaison between client and NSSS vendor.
1974 - 1975 Performed technical and licensability review of PSAR and related sections of the Environmental Report.
Assisted with shielding design experiments conducted at the University of Michigan.
EDUCATION:
B.h.NuclearEngineering,UniversicyofMichigan.
Graduate Engineering Studies, University of Michigan.
SOCIETIES:
American Nuclear Society
District of Columbia) SS.
UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of
)
)
CONSUMERS POWER COMPANY
)
Docket No. 50-155
)
(Big Rock Point Nuclear Power Plant))
AFFIDAVIT OF CIIARLES E.
AXTELL I Charles E.
Axtell, of lawful age, being first duly sworn, do state as follows:
I am employed by Consumers Power Company as the Plant Health Physicist at the Big Rock Point Plant.
I have held this job for 12 yecrs.
In this job my respon-sibilities include monitoring and controlling personnel exposure, ALARA considerations, controlling off-site releases of radioactive materials, and plant water chem-istry.
My resume is attached.
I am primarily responsible for the responses to Interrogatories 7-2, 7-7, and 7-11.
I have joint respon-sibility for the responses to Interrogatories 7-10 and 2-14.
.. To the best of my knowledge and belief, the state-ments in >.his affidavit, the responses to the above inter-rogatories and the information on the data sheets are true and correct.
Cint. s'OA, Ef Charles E'.'IAxtell Subscribed and sworn to before me this 22nd day of February, 1900 4
?l U ' a Y :
?! ' l s'
Notary Public'
- /
My Cununi. ion Eqho Syra'xt 1+,1932
s -
C. E. AXTELL, PLANT HEALTH PHYSICIST Mr. Axtell was assigned as a Senior Chemical Laboratory Technician at the Big Rock Point Nuclear Plant in May 1961.
EDUCATION:
1.
About 2 years of college at Bay City Junior College, Jackson Junior College and North Central Michigan College. Night school, general college.
1958-1973.
2.
International Correspondence School, 1/58 - 1/60 (two years).
Chemistry (completed).
3.
Mathematics Course, Consumers Power Company, General Office, Jackson, Mic".
2n, 5/61 (two weeks).
4.
Jackson Junior College, Jackson, Michigan, 6/61 - 7/61 (eight weeks).
Basic reactor technology; applied reactor physics and health physics; and, use and interpretation of radiation measurement equipment, in-cluding foil irradiation and use of actual monitoring equipment destined for the Big Rock Point Nuclear Plant.
5.
US Public Health Service, Cincinnati, Ohio, 1/62 (two weeks).
Basics of Radiological Health.
Subjects ccvered: major sources of radiation exposure; modes of radiation injury; basic units of termi-nology; standards for radiation protection; and, use, operation and evaluation of monitoring equipment.
6.
US Public Health Service, Cincinnati, Ohio, 2/62 (two weeks).
Pollutants in water; methods and techniques for determination of radiocuelides in the aquatic eu ronment; and, lectures, problem sessions and laboratory practice.
7.
US Public Health Service, Cincinnati, Ohio, 2/62 (one week).
Pollutants in air; methods and techniques for determination of radio-nuclides in the air environment; and, lectures, problem sessions and
- laboratory practice.
8.
US Public Health Service, Austin, Texas, September 9,1968 (two weeks).
Occupational Radiation Protection.
Subjects covered:
beta and ga==a shielding design; disposition of radioactive vastes; neutron instru-mentation; prote: tion and biological effects; transportation accidents and regulations; emergency exposure and concentration guides; emergency planning and management; and evaluation of radiation exposures.
9.
Consumers Power Company, Jackson, Michigan, September 30 - October 4, 1968 (one week).
Basics of Job Management.
Some of the subjects covered:
Company History and Organization; Work Management; Management by Objectives; and Decision Making.
2 EDUCATION (Contd) 10.
Consu=ers Power Co=pany, Jackson, Michigan, Nove=ber 11-15, 1968 (one week).
Basics of Job Management.
Some of the subjects covered:
Equal employment Opportunity Policy; Accounting Records; Rate Making; Union Relations; and Self-Improve =ent.
11.
US Public Health Service, Las Vegas, Nevada, February 24 - March 7, 1969 (two weeks).
Radioauclide Analysis by Gnmma Spectroscopy. The course covered the theory and operation of a gam =a spectrometer; considerations necessary for the selection of a spectrometer; and the procedures for evaluating its performance.
The course also covered the consideration of spectral analysis methods, including hand calculation and computer methods.
12.
US Public Health Service, Winchester, Massachusetts, August 10-14, 1970 (one week).
Management of Radiation Accidents.
Major attention was devoted to potential sources and types of accidents. preplanning first stage management and follow-up and public relations.
Selected accidents were reviewed in detail.
Claas and panel discussions enabled the trainee to discuss specific problems with other class members, the training staff and consultants.
Field exercises were also held.
13.
Special Training Division, Oak Ridge Associated Universities, Oak Ridge, Tennessee, February 23 - May 1, 1976 (een weeks).
The Health Physics Course included:
Radiation Physics, Spectral Analysis, Counting Statistics, Survey Instruments, Radiobiology X-ray Production, Neutron Production, Nuclear Medicine, Advanced Absolute Counting, Liquid Scintillation Counting, Radiation Protection Guides, X-ray Fluorescence, Industrial Hygiene, Neutron Activation, Solid State Spectroscopy, Radia-tion Accidents, Film Dosimetry, Radiophotoluminescent and Ther=olumine-scence Dosimetry, E=ergency Plans and Procedures, Environ = ental Monitoring, Waste Disposal, Air Sampling, Laser Safety, Transuranium Health Physics, Accelerator Health Physics, Microwave and Accident Dosimetry.
14.
Respiratory Protection Programs, conducted by the Los Alamos Scientific Laboratory, Boston, MA, November 3-5,1976.
Air Purifying Respirators, Atmosphere Supplying Respirators, Respiratory Physiology, Oxygen Deficient Ataosphere, Breathing Air Quality, ANSI Standard Z88.2-1969, Minimni Acceptable Respiratory Program, Concept of Protection Factors, Fitting of Respirators, Respiratory Fitting and Training of the Wearer, Medical Surveillance, Inspection and Maintenance, Respirator Use During Emergencies.
PUBLICATIONS 1.
R. W. Sinderman and C. E. Axtell, Personnel Radiation Exuosure Aspects of Ooeration, Maintenance and Refueling a Boiling Water Reactor, p,r_es, ente _d 'at the Health Physics Society Mid-Year Topical Symposium, Los Angeles, Cali-fornia, January 29-31, 1969.
- In chronological order.
3.
PUBLICATIONS (Contd) 2.
C. J. Hartman and C. E. Axtell, Unusual Corrosion Problems Associated with a Boiling Water Reactor, presented before the National Association of Corrosion Engineers, Houston, Texas, March 10-14, 1969.
3.
R. W. Sinder=an and C. E. Axtell, Liould Radioactive Waste Management, presented before the American Institute of Chemical Engineers, Cincinnati, Ohio, May 16-19, 1971.
4.
C. E. Axtell and B. L. Murri, Methods Utilized to Reduce Radioactive Liould Discharge, presented at the Health Physics Society Annual Meeting, Las Vegas, Nevada, June 12-16, 1972.
EXPERIENCE:
1.
9/57 - 5/61, Plant Laboratory Technician, J. C. Weadock Steam Generatich Plant (614 MWe)
Assisted in chemical analysis and control; instrument calibration and repair; and plant efficiency analysis.
2.
9/61 - 10/61 (10 weeks), Vallecitos Atomic Laboratory.
Assigned to the Radiochemistry Section of Vallecitos Atomic Laboratory.
Participated in all phases of radiochemical analysis and control associated with a boiling water reactor.
3.
3/62 - 1/68 Assigned to the Big Rock Point Nuclear Plant as Senior Che=ical and Radiation Protection Technician.
4.
2/68, Big Rock Point Nuclear Plant Assigned the titl4 of Supervisor., Chemical and Radiati.on Protection.
5.
Secretary-Elect, Power Reactor Health Physicist Group, 1972-1973.
6.
Member, Health Physics Society.
7.
Three-week (120 hours0.00139 days <br />0.0333 hours <br />1.984127e-4 weeks <br />4.566e-5 months <br />) course,1975.
Conducted Basic Health Physics Course for six Chemical and Radiation Protection Technicians. The course covered such topics as: Mathematics, Basic Physics, Radioactivity, Radiation and Contamination, Biological Effects, Units, Protection Against Radiation and Contamination, Standards and Guide Values, Detection and Measurement of Radiation and Centamination, Personnel Monitoring, Survey Techniques, Plant Monitoring of Radiation and Contamination, Decontamination, Waste Disposal, Environmental Safety, Emergency Actions, Shielding Calculations, 10CFR19, 10CFR20, Radiation and Respiratory Protection Program, Hazards Associated with Certain Isotopes, Portable Survey Instrument Calibration, Process Monitors, Detector Fundamentals, Technical Specification Requirements, Air Sampling, Shipment of Radioactive Materials, Site Emergency Plan, and Plant Procedures Manual.
8.
3/76 - Assigned the present title of Plant Health Physicist.
District of Columbia) SS.
UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of
)
)
CONSUMERS POWER COMPANY
)
Docket No. 50-155
)
(Big Rock Point Nuclear Power Plant))
AFFIDAVIT OF ROGER W.
SINDERMAN I,
Roger W.
Sinderman, of lawful age, being first duly sworn, do state as follows:
I am employed by Consumers Power Company as a corporate Health Physicist.
In this job I am respon-sible for all corporate matters relating to radiological control.
My resume is attached.
I am primarily responsible for the answers to Interrogatories 7-1, 7-3, 7-4, 7-5, 7-6 and 7-8.
In addi-tion, I have joint responsibility for the response to Interrogatory 2-14.
. To the best of my knowledge and belief, the state-ments in this affidavit and the responses to the above interrogatories are true and correct.
?/
'bcvm,I ]C r
us 2L., e. ~r)
~
Joger W.
Sinddrman Subscribed and sworn to before me this 22nd day of February, 1980 N l! LA-l,,4 'd,'k_bh Notary Public
/
M, Ccmmi.uion Ev7 :a Sq der 14, 1932 7
ROGER WILLIAM SINDERMAN Education:
B.S.
Science Engineering, University of Michigan M.S.
Health Physics, University of Michigan M.P.H.
Health Physics, University of Michigan Experience:
Consumers Power Company as Corporate Health 1974 tc.
Physicist responsible for all aspects of radio-Present logical control at Consumers Power Company nuclear facilities.
These responsibilities include radia-tion exposure to employees, environmental sur-veillance, radioactive waste and effluent control.
1973-1974 Consumers Power Company as Palisades Plant Health (6 month Physicist responsible for radiation r otection, period) effluent and environmental control a-the Palisades Plant.
1971-1973 Consumers Power Company as Environmental Health Physicist responsible for environmental radiological surveillence and control of radiological effluents from the Company's nuclear facilities.
1968-1971 Consumers Power Company as Health Physicist respon-sible for Big Rock Point Plant radiological control and Palisades Plant construction activities related to radiation protection.
1966-1968 Consumers Power Company as Associate Engineer, General Engineer, and Chemical and Radiation Pro-tection Supervisor at the Big Rock Point Plant responsible for Plant radiation protection activ-ities and various engineering tasks.
Societies:
Health Physics Society American Public Health Association
s E
UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING 20ARD In the Matter of
)
)
CONSUMERS POWER COMPANY
)
Docket No. 50-155
)
(Big Rock Point Nuclear
)
Power Plant)
)
CERTIFICATE OF SERVICE I hereby certify that copies of " Answers of Consumers Power Company to Interrogatories Propounded by Christa-Maria, et al.,"
and attached affidavits and refer-enced documents in the above-captioned proceeding were served upon the following persons by depositing copies thereof in the United States mail, first class postage prepaid, this 22th day of February, 1980.
Herbert Grossman, Esq.
Janice E.
Moore, Esq.
Atomic Safety and Licensing Counsel for NRC Staff Board Panel U.S.
Nuclear Regulatory U.S.
Nuclear Regulatory Commission Commission Washington, D.C.
20555 Washington, D.C.
20555 John O'Neill, II Dr. Oscar H.
Paris Route 2, Box 44 Atomic Safety and Licensing Maple City, Michigan 49664 Board Panel U.S.
Nuclear Regulatory Christa-Maria Commission Route 2, Box 108C Washington, D.C.
20555 Charlevoix, Michigan 49720 Mr. Frederick J. Shon Atomic Safety and Licensing Atomic Safety and Licensing Appeal Board Panel Board Panel U.S.
Nuclear Regulatory U.S.
Nuclear Regulatory Commission Commission Washington, D.C.
20555 Washington, D.C.
20555
s
, Atomic Safety and Licensing Board Panel U.S. Nuclear Regulatory Commission Washington, D.C.
20555 Docketing and Service Section U.S.
Nuclear Regulatory Commission Washington, D.C.
20555 Karin P.
Sheldon, Esq.
Sheldon, Harmon & Weiss 1725 I Street, N.W.
Suite 506 Washington, D.C.
20006 m-.
seph/Gallo ne of the Attorneys for Consumers Power Company
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STATDENT By Roger W. Sinderman before the Soecial Joint Committee on Nuclear Energy January 21, 1980 e
1 SPECIAL JOI?iT CC5'MITTEE ON NUCLEAR ENEEGY - JA?iUARY 21, 1980 Mr Chairman, members of the co==ittee, I am Roger W Sinderman, Corporate Health Physicist of Consumers Power Company, Jackson, ?Hehigan.
In a previous appear-ance before this co==ittee, I presented information on low level radioactive vastes resulting from the operations of our nuclear power facilities.
In that testimony, a physical description of ILv level vastes, their bulk radioactivity content, the volumes produced and how and where they were shipped for disposal was described.
Today, I would like to take a broader view and provide detailed analyses of the specific radioactive content of high level vaste, lov level vaste and effluents from our nuclear facilities.
In addition, I will present data showing how the radioactive content of each of these vastes decreases with time to provide a more clear focus on the longevity of any hazard associated with the disposal of each.
I have also taken the liberty of bringing with me a sensitive radiation detection instrument and some radioactive material, including samples of actual lov level radioactive vaste from our Palisades Plant.
Perhaps after my formal remarks are concluded, I can demonstrate and compare the levels of radioactive materials in these vastes to levels of radioactive material in products we all come into conmon contact with.
As you vill see, I have brought along a few of these items also.
Initially, in order to more easily relate and co= pare some of the ter=s I will use, I would like to describe the nature of radioactivity.
First we must recall some fundm=entals of atomic structure. An atom is basically composed of a nucleus surrounded by orbiting electrons.
Most of the nuclei of atoms in this world are stable.
That is, they contain no excess energy.
Some nuclei, however, contain excess energy.
In this state, they are unstable.
The excess energy, and hence instability, is eliminated by the emission of particles or electromagnetic rays from the nucleus.
This loss of excess energy by ejection of a particle or a ray is called radioactive decay.
2 SPECIAL JOINT CCKETTEE ON NUCLEAR ENERGY - JA*iUARY 21, 1980 The a=o"nt of radioactive atc=s in a sample of radioactive material is described in terms of Curies.
A curie is equal to 37 billion unstable atc=s decaying (e=itting =nergy) each second.
On the other hand, the rate at which radioactive material decays is described in terms of its half life - that is, the time it takes for one half the atoms in a particular sa=ple of radioactive material to get rid of their excess energy.
Curiously, it takes an equal a=ount of time -
another half life - to get rid of half of the remaining half while yet another half life is required to reduce the radioactivity to half of the quarter that was left and so on.
(1, 1/2, 1/L, 1/8, 1/16...)
One can quickly see that all of the radioactivity of any radioactive =aterial never cc=pletely disappears under such a scheme of reducing by halves.
- However, at some point, a level of radioactive decay is reached for all radioactive =a-terials including those in radioactive vastes, which, for all practical purposes,
=ay be presumed acceptably safe.
These measures of time vill, of course, vary depending upon the initial amount of radioactivity present; the half lives of the radioactive material involved and the types of particles and rays emitted.
Though various types of particles are involved, the one of principle i=portance in both effluents and vaste from nuclear power facilities is the beta particle.
To a lesser extent, sc=e of the high level vaste - the radioactivity =aterial in spent nuclear fuel - contains alpha particle emitting radioactivity. Accordingly, the ga==a ray and x-ray are the only forms of electromagnetic rays emitted frc=
radioactive material in effluents and vastes of nuclear power facilities.
When these particles and rays interact with matter, they impart their energy to orbiting electrons of ators in matter. This action strips electrons from their orbits creating ions - hence the ter= ionizing radiation.
Of greater importance, however, is the fact that these particles and rays do not interact with the nucleus of atoms in a manner which makes them unstable (radioactive).
3 SPECIAL JOINT C0bMITTEE ON NUCLEAR ENERGY - JANUARY 21, 1980 In other words, being exposed to radiation does not cause or create radioactivity.
(Demonstration same by use of instru=ent and sources of radioactive material).
Exposure to radiation (alpha, beta, ga==a and x-radiation) =erely imparts energy to the recipient.
In contrast to exposure to other forms of electro =agnetic radiation, such as visible light, ionizing radiation has a greater penetrating ability. When exposed to light, our bodies, for example, only absorb the light's energy on the very outer surface. Alpha, beta, ga==a and x-radiation penetrate farther imparting energy within various tissues and organs.
This absorption of raergy and resulting creation of ions causes chemical changes which are the root f all biological effects of radiation exposure. To understand these effects better, I have attached a reprint of an article from the July-August FDA Consuner entitled, " Radiation".
The difference in penetrating ability of various for=s of radiation may be des-cribed as follows:
Alpha particles stopped by a few inches of air or a single sheet of paper Beta particles stopped by several yards of air or a thin sheet, 1/8 inch or less, of steel
~
Ga==a and X-rays essentially stopped by several feet of concrete or a few inches of steel or lead (Demonstration by use of instrunent and sources of radioactive material).
From the above, we can begin to understand that radioactive material produces different kinds of hazards depending on the type of radiation emitted as well as the length of time it takes the particular material to decay to inocuous levels. An alpha emitting material like plutonium, for example, cannot expose any substance kept greater than a few inches from it.
A simple metal, glass or
k SPECIAL JOINT CC'04mt.E ON NUCLEAR E'!ERGY - JANUARY 21, 1980 paper container also solves any exposure problem associated with alpha emitting materials.
Likewise, beta emitting materials such as strontiu=, if contained within containers of relatively co==en structural properties, present no hazard while ga=na and x-ray emitting radioactive material such as Xenon and Krypton require containers of greater thickness or mass to reduce the number of rays penetrating the container valls to permissible levels.
On the other hand, ingestion or inhalation of radioactive =aterials may result in uptake of the =aterial by various tissues and organs.
This material, whether alpha, beta, gn==a or x-ray e=itting, now inside the body, irradiates (imparts energy to) live tissue.
Because alpha particles lose all their energy within a very small thickness of tissue, more ions are created per unit thickness of tissue and hence more damage results than vould be caused by a loss of a similar amount of energy by beta, ga==a or x-radiations. Depending then upon the nature of the concern, exposure from a source external to the body or exposure from a source internally within the body, alpha =ay be more damaging than ga==a or vice versa.
(De=onstration difference between external exposure hazard and contamina-tion hazard).
Let us nov look at the tables. Table I shows the amount of radioactive caterial (high level vaste) in spent nuclear fuel after the fuel has been in the reactor for three years.
The 'inits are in Curies per therral =egawatt. A three year period has been chosen because typically one third of the fuel in a reactor is re=oved during a yearly refueling.
For our Palisades Plant, as an exa=ple, 68 fuel bundles are recoved during each refueling.
These bundles have produced one third of the reactor core's energy over the three years they were in the reactor. At its licensed power level of 25h0 thermal =egawatts, each bundle was responsible for producing 25Lo 3x68 12.5 thermal =egawatts
5 SPECIAL JOINT CCEfITTEE ON NUCLEAR ENERGY - JANUARY T. 1900 Therefore, the numbers in the table can be multiplied by about 12.5 to obtain the radioactivity levels in a single spent fuel bundle for our Palisades Plant.
For Big Rock Point, the corresponding conversion is 2.7 thermal =egawatts per bundle.
The table shows all fission product radioactive material of any consequence plus the i=portant transuranium heavy metal alpha emitting radioactive material produced.
The quantities of these materials as a function of time is also shown for decay periods up to 10,000 years after removal from the reactor.
The quan-tities of fission products shown are constant per thermal =egawatt for all light water nuclear reactors.
The quantities of transuranium =aterials given are for our Palisades facility and =ay vary slightly for other reactors.
It is L:portant to note that when a uraniu= atom splits or fissions, the products of this fission are normally unstable, hence radioactive.
The fissioning event nor= ally produces 2 lighter atoms and additional neutrons which can cause suc-ceeding fissions.
These neutrons, particles from the nucleus, can also be ab-sorbed by other nuclei and result in unstable nuclei, hence cause radioactivity.
This important exception to the point made earlier about exposure to alpha, beta, gn=ma and x-radiation not causing radioactivity to be produced exists only in the reactor core, not in vaste, but gives rise to the transuranium radioactive materials of neptunium, plutonium, americium and so on.
This occurs because of neutron absorption rather than neutron fission of uranium.
Neutron absorption by other =aterials in the reactor also gives rise to several radioactive =aterials in lov level vastes and effluents of our nuclear reactors which I shall describe in more detail later.
As you will note, the initial radioactive content of spent fuel is very high - over h million curies per therm d =egawatt.
This level of radioactive material requires several feet of water or concrete shielding to reduce the external exposure hazard
6 SPECIAL JOINT COMMITTEE ON NUCLEAR ENERGY - JA'IUARY 21, 1980 to per=1ssible levels.
The total energy emitted by the decaying radioactive material is sufficient to cause the fuel to increase in te=perature unless cooling is provided.
Initial decay, however, is quite rapid.
For example, after only one day of decay, the radioactive content is reduced to less than 25%
of the amount initially present at reactor shutdown:
1.e.
952,200 r ab ut 23% of the initial enount present h,196,000 Decay slows considerably
.'ter 1 year as seen in the table with the re=aining radioactive content of 1.8% of the original after 1 year while at 10,000 years, only 0.0003% remains.
After 1000 years, the significant contributers are selenium, sirconium, technetium, and tin, as fission products and plutonium and americium as transuranium scecies.
At this point, the external radiatien exposure hazard is minimal because all re-maining fission and transuranium products are alpha and beta emitters with relatively s=all a=ounts of X-and ga==a radiation involved.
Hence, ingestion and inhalation are the only pathways that may still exist for significant human exposure.
Huwever, this pathvey is only possible if these metals were sc=ehow released to the biosphere in a form easily respirable or ingestable - no small fe at.
However, it may still be useful to compare ingestion of this material, for example, to the ingestion of uranium ore - the natural parent product from which the fuel was fabricated.
First, if we assume that the spent fuel is reprocessed so that the.ransuranium elements are removed and recycled back into nuclear fuel to produce more energy, the ingestien toxicity of the fission products after about 1000 years of decay, pose about the same ingestion hazard as ingesting uranium ore.
Recond, if the spent fuel is not reprocessed, it would take about 150,000 years to reduce the ingestien toxicity to that of uranium ore.
7 SPECIAL JOINT C0'OfI"IEE CN NUCLEAR E'!ERGY - JANUARY 21, 1980 To provide a more complete picture of the radioactive decay involved, I have tabulated in Tables II, III and IV all of the forms (isotopes) of the radioactive material for three of the elements shown in Table I.
First, in Table II, strontium, as an example of a relatively long lived metallic fission product, is shcvn. Next,
xenon, as an example of a relatively short lived gaseous fission product, is shown in Table III.
Finally, in Table IV, I have shown the significant forms of plutonium as an exa=ple of an alpha emitting long lived transuranium material.
In Tables V, VI and VII, radioactive materials contained in lov level vaste from our palisades facility is tabulated.
Though variations can be expected depending upon many operating variables, the values presented here can be taken as typical.
Thr.se values are taken frem our palisades plant vaste products during 1979 As you vill note, many of these materials are isotopes of the same elements shown in Table I for high 1^ vel vaste.
This is true partially because small amounts of uranium circulate in the reactor coolant and as this uranium passes through the neutron field in the reactor core, sc=e of this uranium fissions producing
^hese same fission products.
So.ne fuel cladding may also exhibit small defects,
thus permitting the leakage of scme of the fission products from the fuel directly into the coolant.
The other radicactive materials shown in Tables V, VI and VII are called activation products.
These radioactive materials, cobalt, manganese iron, chromium and sodium are produced when their stable counterparts which exist as trace impurities in the reactor coolant ibsorb neutrons and beccce radioactive.
The reactor coolant requires ecnstant purificat. en, by filtration, demineralization and distillation.
The resulting filters, deminera_izer resins and solidified evaporator sludges contain most of the radioactive material originally in the coolant.
This material is then stored on site until a sufficient quantity is produced to ecenemically ship off site for permanent disposal as described more fully in my previous testimony.
6 SPECI.. JOINT COMMITTEE CN NUCLEAR ETERGY - JANUd.Y 21, 1980 As you vill note from an examination of Tables V and VI, the radioactive materials in low level vaste are cuch less in quantity than high level vaste and in general, exhibit shorter half lives.
Because almost all of these radioactive materials emit ga==a as well as beta radiation, an external hazard exists as well as an ingestion / inhalation hazard. Burial and shielding by only a few feet of soil is sufficient to reduce the external hazard to permissible levels at zero decay time.
The ingestion / inhalation hazard at zero decay time is also lov and is less than the hazard associated with burial of many other substances.
Finally, in Tables VIII and IX, I have tabulated radioactive effluent releases to both the atmosphere and Lake Michigan from our Palisades plant for the period July 1978 through June 1979 Decay periods up to 100 years are shown again pro-viding a comparison to the lengevity of both high level and low level vastes.
These values may be considered typical for large pressurized water reactors though some variation can be expected to occur depending upon specific design parameters at each particular nuclear facility.
As with lov level vastes, you will note a similarity in the radioactive caterials involved and relatively short half lives.
You vill also note that I have separated tritium (the radioactive form o.* hydrogen) from the total.
Eecause it exists as part of the wate. =olecule in the reactor coolant, it cannot be filtered or removed by any of the processing steps of the radioactive vaste system.
Fortunately,
its mode of radioactive decay is by emission of a very weak beta particle and therefore, its hazard relative to the other radioactive materials present is tinimal.
Here the effect of such effluent releases may best be compared with typical exposures to radiation.
These releases expose people residing at the site pro-perty boundary from both the external exposure pathway as well as.he inhalation /
ingection pathway to less than 1% of the radiation exposure they receive from natural background.
People residing at greater distances are, of course, exposed to even lower levels of radiation.
9 SPECIAL JOINT CCM4I'"ITE ON NUCLEAR E'SGY - JANUARY 21, 1960 I hope the above discussion and the tabulated data provides the Co=ittee with the infor=ation it needs to help develop a rational approach to the use of nuclear energy in Michigan.
I would like to thank this Cor.ittee for the opportunity to make this presentation and will gladly provide additional information should you so desire.
O e
Table I Fission and Transuranium Raiioactive Material in Spent Puel As a Function of Decay Time in Curies Per "her ::a1 Megavatt' 100 1000 10,0C0
" "TU'S Shutdev.
1 Hour i Day 10 Oays 1 Menth 1 Year 10 Years Years Years Ye ars line 2
Callium 97 cer=anium 296 Arsenic k,617 Selenium 21,730 L65
.9
.03
.01
.01
.01
.01
.01
.01 3romine 61,080 3,566 Krypton 101,100 28,270
~
Rubidium 168,600 1,331,000 5,003 Strontium 19k,700 76,590 30,030 22,150 17,430 2.559 1,921 202 Y1trium 283,600 12L,500 Lh,960 30,020 24,2Lo 2,83L 1,921 253 lireenium 176,600 90.830 63,210 41,850 33,810 95k
.10
.10
.10
.10 Niobium 33L,500 138,500 83,080 47,520 hk,350 2,050 Molybdenum 22L,900 53,910 39,270 h,063 Technetiu=
298,900 89,000 37,780 3,93k 26 5
5 5
5 5
Ruthenium 130,200 90,910 59,990 52,970 L2,160 9.6L8 19 Rhodium 168.100 106,200 80,850 52,670 bl,700 9,6L7 19 Palladiu=
11 h50 8,523 2.725 Silver ik,700 10,1LO h,225 Cadmium 1,!. 30 Indium 232 Tin 36,610 5,123 61h 262 93 7
.05
.03
.01
.01 Antimeny 117.100 2k,850 3,LTS 1,101 h92 Tellurium 202,300 99.130 h3,570 8,168 2,103 IL2 Iodine 313,800 181. LOO 56,L60 12.550 2.222 Xenon 2L3.100 78,690 61.L90 17,020 1,265 Cesium 207, LOO 788, LOO 6,369 6,026 5,6L2 L,892 2,90h 355 Barium 235, LOO 112, LOO LB.150 30,860 12,650 3,266 2,653 332 Lanthanum 216,000 127.300 L9,L90 31,600 10,720 Cerium 196.000 126,600 109,200 7k,650 58,770 IL,L60 Praseedymiu= 158,200 113,900 78,710 62,080 L2,710 14 hko Neodymium 32,560 24.010 15,800 9,007 2,563 Promethium 35,580 Jo,030 23.810 8,787 7,096 5,288 L9 Sa=arium 9,632 8,270 5,718 Europium 32 cadolinium 169 Terbium 60 Dysprosium 7
Holmium 5
Tetal Fissien k,191,000 3,866,000 9k7, Loo 520, Loo 350,600 70,700 9,610 1,1L5
.6
.6 Neptunium
.008
.008
.008
.008
.008
.008
.008
.008
.008
.008 Plutonium h,689 h,689 L,689 h,689 h,689 4,689 2,950 8k 2k 13 Americium 6
6 6
6 6
6 6
5 2
3 Curium 101 101 101 101 101 101 71 2
Total Transuranic L,706 L,796 h,796 h,706 L,796 L,790 3,030
_91
_26
_13 Total h,196.000 3,871,000 952,200 525,200 355, LOO T5,500 12,6ho 1,236 27 1h "Cak Pidge National Laboratory CRICE3 Co=puter Code for Decay times to 1 year for fission products. All others personally calculated by the writer.
Table II Radioactive Forms of Strontium in Spent Fuel As a Punction of Decay Time in Curies per Thermal Megawatt
- 100 1000 10,000 ISOTOPES Shutdown 1 Min 1 Ifour 1 Day 10 Days 1 Month 1 Year 10 Years Years Years Years Strontium 86 88 89 22,h80 22,h80 22,560 22,270 19,690 14,970 153 90 2,h64 2,h64 2,46's 2,h6h 2,h63 2,h60 2,406 1,921 202 91 28,660 28,810 27,080 5,234
.001 92 31,110 31,200 24,2ho 66 93 37,200 35,110 254 94 35,780 22,230 95 33,140 6,712 97 3.817
.003 Total 194,700 149,000 76,590 30,030 22,150 17,430 2,559 1,921 202 "Onk Ridge National Laboratory ORIGEN Computer Code for Decay times to 1 year for fission products. All others personally calculated by the writer.
TABLE III Radioactive Forms of Xenon in Spent Fuel As a Function of Decay Time in Curies per Thermal Megawatt
- 100 1000 10,000 ISOTOPES Shutdown 1 Min 1 Hour 1 Day 10 Days 1 Month 1 Year 10 Years Years
, ars Years Xenon 128 129 130 131 176 176 176 171 123 49 131 132 133 1,95h 1,954 1,937 1,5h7 113 3
133 56,h80 56,h80 56,310 51,710 16,790 1,215 134 135 16,980 16,530 6,957 669 135 9,781 9,812 11,0ho 7,388
.001 136 137 h7,050 40,150
.8 138 hh,330 h2,260 2,277 139 39,510 lb,500 1h0 19,370 929 1h1 5,901 1h2 1,321 1h3 195 1hh 21
.1 Total 2h3,100 182,800 78,690 61,490 17,020 1,265
" Oak Ridge National Laboratory ORIGEN Computer Code for Decay times to 1 year for fission products.
All others personally calculated by the writer.
Table IV Itadioactive Forms of Plutonium in Spent Fuel As a Function of Decay Time in Curies per Thermal Megawatt" 100 100 10,000 ISOTOPES Shutdown 1 Min 1liour 1 Day 10 Days 1 Month 1 Year 10 Years Years Years Years Plutonium 238 73 73 73 73 73 73 73 68 33 239 9
9 9
9 9
9 9
9 9
9 7
240 17 17 17 17 17 17 17 17 17 15 6
241 h,590 4,590 4,590 h,590 4,590 h,590 h,580 2,860 25 242
.08
.08
.08
.08
.08
.08
.08
.08
.08
.08
.08 Total 4,689 4,689 h,689 h,689 4,689 4,689 h,689 2,950 84 2h 13
- Personally calculated by the writer.
Table V Radioactive Waste in compressible Low-Level Waste Isotopes llalf Annual Off-Site Amount Remaining After Li fe Shipments 1 Month 1 Year 100 Year.
Curies Manganese 54 303d
.201
.188
.087 cobalt 58 71.3d 92
.69
.029 Cobalt 60 5.263y 5^
52
.h6 cesium 13h 2.04y
.069
.067
.049 cesium 137 30.0y
.26
.25
.2h
.02h Total 1 98 1 715
.865
.024
Table VI Radioactive Waste in Spent Demineralized Resins Annual Off-Site Amount Remaining After Isotopes IIalf Life Shipments 1 Month 1 Year 100 Years Curies Manganese 54 303d 5.36 5
2.32 Cobalt 57 270d
.0301
.0279
.0118 Cobalt 58 71.3d 8.59 6.47
.269 Cobalt 60 5.263y 1 90 1.88 1.66 Niobium 95 35
.042
.0235 Cesium 134 2.0hy 2.68 2.61 1 91 Cesium 137 30.0y 5.91 5.90 5.78 59 Total 24.5 21.9 11.95
.59
Table VII Radioactive Waste in Solidified Evaporator Concentrate Annual Off-Site Amount Remaining After Isotopes Half Life Shipments 1 Month 1 Year 100 Years Curies Manganese Sh 303d 2.22 2.07 962 Cobalt 57 270d
.0397
.0368
.0156 Cobalt 58 71.3d 2.81 2.11
.088 Cobalt 60 5.263y 2.10 2.08 1.8h Cesium 13h 2.0hy 3.27 3.18 2.33 Cesium 137 30.0y 1h 1h 13.7 1.39 Cerium lhh 28hd
.070
.0652
.0288 Total 24.5 23.5 18.96 1.39
Table VIII Atmospheric Releases of Radioactive Material from Nuclear Power Facilities" Isotopes Half Annual Amount Remaining After Released Life Release 1 Month 1 Year 100 Years Mil 11 curies Krypton 85 10 76y 2,590 2,580 2,h30 h.13 Xenon 133 5.27d 156,960 3,037 Xenon 135 9.1hh 1,675 Xenon 131m 11.8d 2,h9h h28 Xenon 133m 2.26d 1,332 0.13 Krypton 87 76m 288 Krypton 85m h.hh 448 Krypton 88 2.8h 736 Argon 41 1.8h 96 Iodine 131 8.05d 19
- 1. bl4 Iodine 133 20.3h 9
Iodine 135 6.6h 3
Cesium 134 2.0hy
.036
.035
.026 Cesium 137 30.0y
.185
.185
.181
.018 Manganese 514 303d
.6308
.589
.27h Cobalt 58 71.3d 2.h3 1.82
.070 Cobalt 60 5.263y 1.81 1.79 1.59 Cobalt 57 270d
.00891
.008
.003 Zirconium 95 6hd
.0752
.054
.001 Niobium 95 35d
.154
.085 Iron 59 45d
.0771
.0h8 Chromium 51 27.7d 1.7
.803 Sodium 214 15h
.02h 1.15 Total 167,597 6,052 2,h32 4
- Actual releases from the Palisades Plant July,1978 through June,1979
Table IX Radioactive Liquid Effluents from Nuclear Power Facilities
- Isotopes Half Actual Amount Remaining After Released Life Release 1 Month 1 Year 100 Years M1111 curies Tritium 12.3y 11,310,000 11,260,000 10,691,000 40,h20 Cesium 134 2.0hy 2
1 94 1.h2 Cesium 137 30.0y 11 11 10 7 19 Iodine 131 8.05d
.058
.004 Cobalt 58 71.3d 2h 17.9
.69 Cobalt 60 5.263y 12 11.9 10 5 kanganese 54 303d 7.8 7.3 3.h Chromium 51 27.7d h.7 2.2 Zirconium 95 6hd
.126
.091
.002 Cobalt 57 270d
.132
.122
.052 Iron 59 h5d 585 369
.002 Lanthanum lh0 40.2h
.852 Xenon 133 5.27d
.187
.00h Total (less Tritium) 63.4 52.8 26.8 1.09 Total 11,31n,000 11,260,000 10,691,000 h0,h20
- Actual releases from the Pc isades Plant July,1978 through June,1979
Table I Fission and Transuranium Radianctive Material in Spent Fuel As a Function of Decay Time in Curies Per Thermal Megawatt
- 100 1000 10,000 ELEMENTS Shutdown 1 Hour 1 Day 10 Days 1 Month 1 Year 10 Years Years Years Years
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6.8.4 Heating and Ventilating i
- 6. 8. 4.1 Heating steam is supplied to various parts of the plant by a 15 psig oil fired package boiler.
6.8.4.2 During the summer, cooling is provided by ventilating through local cooling coils supplied from the service water system.
^
6.8.4.3 The reactor containment vessel and portions of the turbine building are provided with both forced and induced draft vendladon.
This provides for ventilating all potentially contaminated areas so as to avoid the existence of any hazardous condition to the plant proper or surrounding areas during normal plant operadon.
6.8.4.4 The reactor containment vessel ventilation air inlet and outlet pene-trations are each provided with a pair of pneumatically operated valves which close automatically on loss of_poyter o
or on any scram signal. Each pair of valves consists of one swing tfpe vhalve with a high temperature synthetic rubber seat and one butterfly type valve with high temperature
('
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{
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(
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Fuel Damage While Refueling - E>T 3.4, Rev 3 12/9/77 EFT 3.4 FUEL HANDLING ACCIDENTS This procedure is not intended for extended use during recovery or o
restoration following extensive damage incidents wl.ere special equip-ment and procedures will be required.
A fuel handling accident may consisc of the following incidents:
a.
A rapid insertion or drop of fuel into the core.
b.
Transfer cask drop into the core (with fuel in core).
c.
Loss of coolant from the transfer cask during fuel transfer.
d.
Fuel drop into the pool.
e.
Fuel sipping error.
EMP 3.4.1 SYMPTONS a.
Fuel handling cable breaks.
b.
Rapid increase of neutron monitors.
c.
Rapid increase of area monitor readings.
d.
Area monitor alarms.
e.
Neutron monitor alarms.
f.
Reactor scram annunciators, g.
Transfer cask red low-level light is on.
l EMP 3.4.2 AUTOMATIC ACTIONS a.
Reactor safety system may trip if incident occurs over the reactor vessel.
b.
The containment ventilation system will be isolated automatically in the event of an area monitor high radiation alarm from either of the reactor deck area monitors.
EIT 3.4.3 IMMEDIATE OPERATOR ACTION EMP 3.4.3.1.
For rapid fuel insertion into reactor core:
a.
All personnel immediately vacate the sphere.
b.
Initiate a reactor scram.
EMP 3.4.3.2 For transfer cask drop into core:
a.
All personnel immediately vacate the sphere.
b.
Initiate reactor scram.
c.
Inject liquid poison.
Ele 3.4.3.3 For loss of coolant during fuel transfer:
a.
Connect hose from demin water line to transfer cask and supply water to maintain coolant level in cask.
L 2
Fuel Damage While Refueling - EMP 3.h, Rev 3 12/9/77 r
EMP 3.4.3.4 For fuel bundle drop into fuel pool:
4 i
X All personnel i= mediately vacate the sphere.
a.
b.
Close ventilation valves.
E:e 3.h. 3 5 For fuel sipping error:
a.
Flood sipper can with water.
b.
All personnel vacate the sphere.
c.
Close ventilation valves.
EMP 3.k.h SUBSEQUEIT OPERATOR ACTIOUS De 3.h.h.1 For fuel insertion into core:
a.
Initiate Site Energency Plan.
b.
Check reactor scra==ed.
c.
Check ventilation valves closed.
d.
Maintain reactor water level at top of over-flow pipes.
e.
Monitor radiation levels in sphere.
EMP 3.h.h.2 For transfer cask drop into core:
a.
Initiate Site Energency Plan.
b.
Check reactor scrammed.
c.
Check ventilation valves closed.
d.
Maintain reactor vater level below top of overflow pipes.
e.
Monitor radiation levels in sphere.
f.
Re=ove clean-up system from service.
De 3.h.h.3 For loss of coolant during fuel transfer:
a.
When power is restored, nove cask to pool and replace bundle in the storage rack.
HE 3.k.k.h For fuel drop into fuel pool:
a.
Check ventilation valves closed.
b.
Monitor radiation levels in sphere.
EMP 3.h.h.5 For fuel sipping error:
a.
Check ventilation valves closed.
b.
Monitor radiation levels in sphere.
(
,w' m
m SECTION 3 i
m CONTAINMENT m
3.1 GENERAL j
Reactor containment is provided by a spherical steel vessel, 3.1.1 130 feet in diameter. The sphere extends 27 feet below grade and 103 feet above grade. Construction requirements are shown in Drawing C-101. Minimum distance from this vessel to the i
land boundaries of the site is one-half mile, and to the edge of Lake Michigan is 200 feet.
l 3.1. 2 The containment vessel's primary purpose is to prevent a harmful spread of radioactive material to the environi in the event I
of a rupture in the reactor system, or other accident.
To accomplish this end the vessel is designed to withstand the l
internal pressure that would result from the most severe rup-It ture accident which can reasonably be considered possible.
is built to contain this pressure with the greatest practical de-gree of leak-tightnes s.
As a secondary, everyday function, the containnient vess f.
.~
l s
3.1. 3 also serves as a weatherproof housing for the stearn generating system and auxiliaries. Besides the reactor, this includes the steam drum, recirculation piping and pumps, reactor clean-7 up system, shutdown cooling. system, liquid poison system, eme rgency cooling system, and storage and handling facilities for new and j
spent fuel. Figure 3.1 is a cutaway perspective showing the general arrangement inside the sphere.
3 i
The plant is designed so that operating personnel may enter the h1 3.1. 4 sphere and remait. inside as necessary during normal
'~
operation, shutdown and refueling.
a 3.2 DESiCN CRITERIA
~
At an early stage in the design of the plant it was necessary to fix
~
R
- 3. 2.1 the design pressure of the containment vessel in order to pro-A value of 27 psig was conservatively ceed with procurement.
chosen in order to accommodate possible increases in reactor j
The final calcu-system volume during the course of design.
lated peak pressure in the containment is 23 psig, based on the v
u
- s, e
7 Section 3 Page 3 m
assumption of a nearly instantaneous, complete severance of a recirculating pump discharge line, with the reactor in the hot standby condition at 1500 psia. At this time the reactor system contains its maximum stored energy. The calculation p
further assumes the release of all pressurized hot water and steam within the reactor, steam drum, recirculation and cleanup loops, and the steam and feedwater piping to the iso-
~
1ation valves. The containment pressure transient is shown in the following Figure 3. 2.
3.2.2 Design parameters for the containment vessel are as foll3ws:
TABLE 3.1 Design Pres sure, Internal 27 psig Design Pres sure, External 0.5 psig*
'(Coincident with dead load only)
Design Temperature Rise 190 degrees F**
(Coincident with design internal pressure)
Design Maximum Temperature 235 degrees F s
Wind load ASA Std A58.1 Without Snow Load (Basic wind pressure =
30 ps0 s
With Snow Load 60 mph Snow Load ASA Std A58.1 (max = 40 psf at top)
Lateral Seismic Force 5 per cent of gravity (Coincident with dead load and snow load only)
Maximum Leakage Rate at 27 psig
- 0. 5 per cent per day
- External pressure does not govern; with shell thickness designed to withstand 27 psig internal pressure, safe external pressure coincident with dead load only is 1. 22 psig.
- This value assumes a rise from an initial shell temperature of 45 de-grees F, and is structurally more severe than a rise from 100 to 235 degrees F, which is assumed in determining the design maximum temperature. The maximum temperature ris e is used in determining d
secondary stresses due to the structural discontinuity where the ves-sel shell emerges from the foundation. These stresses, when com-bined with primary stresses, are required to be no greater than 1. 5 times the allowable primary stresses.
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UNITED STATES g
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NUCLEAR REGULATORY COMMISSION
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December 21, 1979 Docket Nos. 50-155/255 Mr. David P. Hoffman Nuclear Licensing Administrator Consumers Power Company 212 West Michigan Avenue Jackson, Michigan 49201
Dear Mr. Hoffman:
RE: STATUS AND CATEGORIZATION OF SYSTEMATIC EVALUATION PROGRAM (SEP) TOPICS - BIG ROCK POINT AND PALISADES PLANT Enclosed are tabulations of the status and categorization of the SEP Topics for your facilities. The tabulations indicate those topics for which the initial assessment has been completed, a revised assessment is under way, or the topic is completed.
For these topics we would expect little further involvement from you. For completeness, we have also indicated those topics which are not applicable to your facilities and those generic topics which will
~
be evaluated outside the SEP.
The tabulations also provide a generalized assessment of the expected involvement of the licensee on each topic review. These assessments may change, however, as topic reviews progress and deviations from current licensing criteria are identified.
For those topics requiring a major involvement by you, we will identify as soon as possible the work expected of you. We would then expect you to assign the appropriate resoun:es to these areas so an acceptable review schedule can be met.
Si ncerely,
h-t R chard H. Vollmer, Acting Assistant Director for Systematic Evaluation P rogram Division of Operating Reactors
Enclosure:
Status of cc w/encio See next p DUPLICATE DOCUMENT LI h h ~
4 Entire document previously entered into system under:
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