ML20031D590
| ML20031D590 | |
| Person / Time | |
|---|---|
| Site: | Big Rock Point File:Consumers Energy icon.png |
| Issue date: | 10/02/1981 |
| From: | Blanchard D CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.) |
| To: | |
| Shared Package | |
| ML20031D553 | List: |
| References | |
| ISSUANCES-OLA, NUDOCS 8110130514 | |
| Download: ML20031D590 (22) | |
Text
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1 TESTIMONY OF DAVID P. BLANCHARD CONCERNING CHRISTA-MARIA CONTENTION 8 AND O'NEILL CONTENTION IIIE-2 i
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Christa-Maria Contention 8 and O'Neill Contention IIIE-2 states:
The occurrence of an accident l
similar to TMI-2 which would pre-vent ingress to the containment building for an extended period
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of time would render it impos-sible to maintain the e:panded spent fuel pool in a safe con-i dition and would result in a significantly greater risk to the public health and safety j
than would be the case if the j
increased storage were not allowed.
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I will begin addressing this Contention by assuming i
an accident has occurred which has resulted in the long-term i
uninhabitability of the containment building.
For the 1
purposes of this discussion, the accident which is assumed i
to occur will be a loss-of-coolant accident which results in 4
i i
core damage.
This is an accident similar to that which oc-i curred at TMI-2.
The manner in which this accident prevents i
i access to the containment will be addressed, and the effects-i of the accident on fuei pool equipment and water inventory i
j will be descrit2d.
The consequences of not maintaining the-i-
water level in the fuel pool will be discussed, and I will i
conclude with'a description of the plant modifications
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. proposed to allow remotely-operated fuel pool water inventory i
-maintenance.
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8110130514 811005 PDR ADOCK 05000155 T:
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At the initiation of the loss-of-coolant accident,
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steam will begin entering the containment atmosphere, pres-surizing the containment and elevating the temperature.
At 1.5 psig, a reactor trip wall occur, and all the containment isolation valves will automatically close.
At 2.2 psig, i
i containment sprays will be activated automatically to cool-
.the containment atmosphere.
The water inventory in the primary coolant system will begin to decrease if makeup from normal water sources is not sufficient.
At 4 inches below 4
normal operating level (center line of the steam drum), an evacuation alarm will sound in the containment building; at 8 inches below normal, an additional signal to trip the reactor will occur; and at 17 inches below normal, a low drum level signal will be sent to the Reactor Depressuriza-i tion System, starting a two-minute timer.
As water level continues to fall, it will reach the low reactor level i
j setpoint at 2 feet, 9 inches, above the core.
Still another signal will be generated, insuring the trip of the reactor and-closure of containment isolation valves.
The low reactor water level signal coincidental with low steam drum level, and the timing out of the two-minute timer which was I-
. actuated on low drum level, will result in the opening of valves in four 6-inch lines of the Reactor Depressurization System, quickly depressurizing the reactor and blowing most of tha remainder of the reactor coolant inventory into 1
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containment.
The low reactor pressure will allow the in-jection of cooling water through the low pressure emergency core cooling system.
The maximum containment pressure attainable under these conditions is 23 psig at a temper-ature of 224 degrees F.
Within hours, containment spray will reduce these values to near ambient.
As core spray and containment sprays continue to operate, containment will eventually fill to a level above voich water should no longer be added.
This level is at an elevation approxi-mately 14 feet above the bottom of the containment building.
4 Water is not permitted to rise above this level to prevent j
stressing the containment shell due to the static pressure head of the water.
At this point, water is drawn off the bottom of the containment, pumped through a heat exchanger to cool it, and returned to containment to be injected back into the core through emergency core cooling lines.
In this 2
manner, adequate core cooling and an acceptable containment I
water level can be maintained.
Access to the containment will not be possible during the initial stages of the accident.
The pressure buildup early in the accident is such that the doors on the airlocks normally used to enter containment cannot be opened.
The elevated temperature and hot steem in the atmosphere within the containment make entry hazardous from a personnel i
safety standpoint.
Radioisotopes produced.in the reactor O
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coolant during normal power operation also will be released to the containment during the course of the reactor blow-down.
The amount of these radionuclides, as well as the relatively high rate of decay of some of them, make them unimportant from a radiological standpoint.
By themselves, they do not contribute to making the radiation levels in containment prohibitive.
If core cooling systems perform as designed and prevent fuel cladding failure, access to the containment would be possible soon after the containment sprays condensed all the steam and returned centainment building pressure to normal, perhaps as soon as a few hours after the initiation of the incident.
If fuel cladding failure resulted from this accident and radiation levels became prohibitive, reentry to containment would be limited by plant procedures.
These procedures are developed to be in conformance with NRC standards for radiation protection.
Time would be required to allow cleanup of the containment atmosphere and radionuclide decay.
The delay in attaining access would be dependent on the degree of fuel failure and the amount of fission products released to the containment atmosphere.
If fuel damage involved most of the core as it did at TMI, containment would not become accessible for several months.
Throughout this accident, there is the potential for fuel pool equipment to become affected by the hot steam
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cO and water environment associated with the accident.
To i
i explain what equipment is subject to this environment, I will refer to a simple line diagram (see Figura 1 attached hereto) of the fuel pool and equipment associated with it.
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The diagram shows the location of the containment shell, indicating which of this (,quipment is located inside con-l tainment.
Major hardware associated with the fuel pool i
cooling loop includes the surge tank, filter sock tank, fuel pit pumps, and fuel pit heat exchangers.
Important in removing heat from the fuel pit heat exchangers are the l
component cooling water pumps and heat exchangers.
The remaining equipment on this diagram includes systemc asso-ciated with makeup to the pool which include denineralized water, water from the plant's radwaste system, and fire j
protection system water.
The ability to remove decay heat from the fuel I
pool water is assumed to be lost early during any loss-of-coolant accident including one in/olving a degraded core condition.
This assumption is made because none of the I
equipment associated with fuel pool cooling has been tested 1
i or evaluated in the high_ temperature and high humidity i
environment associated with the LOCA.
This cquipment may,
- in fact, work in such an' environment..The fact that this has not been demonstrated, however, requires us-to assume
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. O conservatively that this equipment fails.
Moreover, be-cause the component cooling water and fuel pit pumps are located below the maximum permissible water level in con-tainment, they may become submerged and thereby rendered inoperative.
Assuming, therefore, that the spent fuel pool cooling equipment will fail at some stage during a LOCA, the water remaining in the pool will begin heating up as a result of the heat still being generated by the spent fuel.
Heatup to the point of boiling may occur, lowering the amount of water in the pool.
If the level drops below the fuel stored in the pool, heating of the fuel rods may occur, followed by cladding failure and a release of addi-tional fission products to the containment.
Providing makeup water to the pool as it boils off can prevent damage to the fuel pool fuel assemblies by keeping them covered with cooling water.
As explained in Mr. Sacramo's testi-mony, the stresses resulting from the thermal gradients due to pool boiling will not adversely affect the integrity of the racks, pool liner, and pool walls and floor.
Consumers Power Company has analyzed and deter-l l
mined the time within which makeup water must be added in a
order to avoid uncovering the spent fuel in the pool.
The pool is 20 feet by 26 feet, with approximately 22 feet of O
water above the fuel.
In addition, this analysis took into account the maximum decay heat that is likely to occur by assuming that the LOCA occurs one month following the pre-vious shutdown.
Thus, the pool was assumed to be filled to its proposed capacity of 441 assemblies with 25 assemblies discharged during the last refueling and 416 assemblies having been placed in the pool from previous refuelings.
As a matter of practicality, one month is the shortest period of time in which refuelings can take place at Big Rock Point.
25 assemblies is the number of assemblies typically dis-charged during each of these refuelings.
Larger fuel pool decay heat loads can result by discharging the full core to the pool.
However, a reactor accident as described in this contention cannot take place with the core discharged to the fuel pool.
It was further assumed that there are no heat losses from the pool other than by vaporization of the water, and that no additional water is supplied to the pool during boiling.
These last two assumptions are conservative in that three of the four walls of the pool interface with the containment atmosphere, and natural heat conduction through these wal.1.s will remove heat and reduce the rate of boiling, perhaps to the point where no boiling will occur at all.
Also, containment sprays are located around the steam l
drum cavity approximately 25 feet above the southwest corner l
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- O of the pool.
Actuation of these sprays during a LOCA, and the condensation of steam which results, provides makeup to the pool for which no credit is being taken.
In order to determine the rate at which water in the pool would be boiled off, the foregoing parameters were used in conjunction with the decay heat generation rate from American Nuclear Society Standard 5.1.
The rate at which water would be boiled off from the pool is 2 gpm, which agrees well with the Staff's analysis presented in Section 3.2.1 of the SER.
The amount of time required to boil off all the water above the fuel is approximately one month.
The calculation follows:
3 Volume of water = 20 x 26 x 22 = 11440 ft 5
= 7.13 x 10 lbm
= 85,570 gallons Heat generation one month after previous refueling 6
shutdown (Ref.: ANS-5.1) = 1.14 x 10 Btu /hr 5
Tine to boil (7.13 x 10 lbm) (1122 Btu /lbm)
=
rs.
all water 6
1.14 10 Btu /hr 85,570 gallons Boiling Rate =
= 2 gpm.
700 hrs.60{-
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The three makeup systems described earlier require entry into containment to open hand-operated valves, allowing 1
water to flow to the pool.
If an accident resulting in core damage were to disable the fuel pool cooling system and render the containment uninhabitable for a period beyond a month, then makeup could not be provided before fuel un-covery occurred.
The release of fission products from the fuel pool to the containment atmosphere could result.
Consumers Power Company has investigated ways to prevent the spent fuel from becoming uncovered in the event of prolonged uninhabitability of containment.
Among the approaches considered were performing tests to demonstrate that heat losses through the pool walls were great enough to prevent boiling, adding remote actuation of valves in the demineralized water system to the pool, adding remotely actuated makeup to the pool from the fire protection system.
The latter modification was found to be the most effective approach of the three.
A line diagram is attached (see Figure 2) to describe the way in which this modification functions.
The diagram depicts the containment sphere with the reactor vessel and fuel pool inside.
Portions of the core spray, containment spray, and fire protection piping are shown.
The dotted line represents the remotely vperated makeup line to the fuel pool.
You will recall from my earlier discussion of the accident that after the reactor has depressurized, the core sprays begin injecting cooling v
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water to the vessel.
The path through which this injection occurs is from the fire pumps, through the yard piping and the valves identified as valves VFP-29 and VFP-30, all located.outside the containment.
Injection to the vessel is then -through the motor-operated valves inside containment, designated MO-7070, 7071, 7051, and 7061.
Containment sprays are actuated by opening motor-operated valves MO-7064 or 7068.
It is important early during the accident to inject as much water as possible into the vessel to cool the For this reason, check valve VPI-300 is used to core.
prevent diversion of core spray water to the fuel pool during this stage of the accident.
High containment water level most likely will be attained within a few days follow-ing the accident.
To prevent further addition of water to the containment, a core spray pump is started from the 4
control room and valves VFP-29 and 30 are closed in the plant's machine shop.
Water is drawn from the bottom of containment, is pumped through the core spray heat exchanger, and flows back to the vessel through valves MO-7070, 7071, 7051, and 7061.
At this time, water makeup to the pool will begin automatically through the new makeup line.
The new line is sized to deliver more than 2 gpm to the pool with the core spray system in service.
The core spray pumps deliver approximately 300 gpm to the vessel.
Diversion of
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the small flow to the pool at this time is not important due i
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to the significantly reduced decay heat load in the core.
Our analysis demonstrates that a makeup capacity of 2 gpm is sufficient to maintain the water level above that needed to cover the spent fuel in the pool during the postulated accident.
r CONCLUSION I have assumed, for purposes of addressing this 3
Contention, that a loss of fuel pool cooling equipment can occur after an accident which renders the containment uninhabitable.
To mitigate the effects of suct ccident l
on the fuel in the pool, a makeup line to the fuel pool has d
been installed which is capable of making up for all pool l
l water losses resulting from boiling due to decay heat.
The
..akeup flow is adequate to account for boiling losses, I
assuming the pool is filled to capacity.
The makeup line is i
j u part of the emergency core cooling system.
It functions in such a way as not to jecpardize core cooling functions early in an accident or result in the addition of an unac-l ceptable amount of water to containment late in an accident.
4 Makeup is capable of being actuated remotely from outside l-the containment.
Because tne analysis for the remote #
makeup system takes account of the additional spent fuel resulting from the proposed' expansion of the spent fuel pool, such expansion imposes no additional risk to the health and' safety of the public.
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L _,l O'NEILL CONTENTION IIB A.
THE CONTENTION The Licensee's plan is deficient in failing to dis-cuss the environmental hazards associated with small to medium leaks of radioactive water from the expanded spent fuel pool.
B.
MATERIAL FACTS AS TO WHICH THERE IS NO GENUINE ISSUE TO BE HEARD 1.
The only potential leakage paths for small to medium leaks of spent fuel pool water are those from the pool itself and those from three connected piping p.
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systems:
the Spent Fuel Pool Cooling System piping,
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the Demineralized Water System piping, and the Treated Waste piping.
(Bordine Testimony,pp. 1, 4).
2.
Small to medium leaks from the spent fuel pool it-self could result only from a puncture of the spent fuel pool liner.
(Bordine Testimony, p.2).
3.
Leaked water from a liner puncture would seep i$to the monitoring trenches, flow through the drain tubes imbedded in the pool floor and collect at one or more of eight manually-operated valves lo-cated beneath the pool floor.
(Bordine Testimony, p.3).
4.
Manual opening of the 3/8" valves at the bottom
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of the drain tubes would cause any leaked water held there to empty into the sample basin and flow thence into the containment sump.
(Bordine Testimoay, p. 3).
S.
The only other possible direct leakage path from the pool is condensation or seepage through the concrete pool floor following a liner puncture.
(Bordine Testimony, p.3).
6.
Any wat r seeping through the concrete pool floor would be collected in the floor drains in the loom d
below the pool and flow to the containment sump.
(Pordine Testinony, p.4).
7.
Water leaked directly from the pool via liner punc-ture therefore could not escape to the environment.
(Bordine Testimony, p.4).
8.
The spent fuel pool cooling system piping is located entirely within containment.
(Bordine Testimony, p.5).
9.
g pipe failure in the SFPCS will not result in the drainage of the spent fuel pool below the concrete weir, the normal water level, nor in the sipnoning of pool water.
(Bordine Testimony, p.5).
10.
The maximum amount of water that could be released by a pipe failure in the SFPCS would be approxi-
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3-O mately 6,000 gallons.
(Bordine Testimony, p.6).
11.
Any leaked water would be collected by the floor drains and thence flow to the enclosure sump and the waste receiver tanks.
(Bordine Testimony, p.6).
12.
The demineralized water system and the treated waste system are connected with the spent fuel pool only indirectly, by virtue of having one con-1 nection each with the SFPCS piping.
(Bordine Testi-mony, p.6).
13.
A pipe failure in either the demineralized water system or the treated waste system could not result 1
i in a release of spent fuel pool water to the en-O vironment because water in both of these systems is flowing into containment, not out of it.
(Bor-dine Testimony, p.6).
14.
No leakage as a result of misoperation is possible at the connection.between the demineralized water system and the SFPCS because the valves at this connection have been placed under lock control.
l Bordine Testimony, p.7).
15.
Any leakage that might occur at the connection be-tween the treated waste system and the SFPCS would j
not be released to the environment because treated I
wastewater is radioactive and is handled under 4
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procedures that reflect the requirements of 10 CFR Part 20 and Part 100.
(Bordine Testimony, 4
p.8).
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16.
Any leakage from the Reactor and Fuel Pit Drain Valve CV/4027 would flow directly to the liquid i
radwaste system.
(Bordine Testimony, p.9).
l 17.
Water leaked from the spent fuel pool through con-i nected piping systems because of pipe or valve failure therefore could not be releared to the environment.
(Bordine Testimony, p.9).
C.
DISCUSSION
()
Mr. O'Neill's contention assumes that small to medium i
leaks from the spent fuel pool or systems connected with it can result in uncontrolled releases of pool water to the environment.
The analysis oi all possible leakage paths from the pool, however, undertaken in the testimony of Thomas C. Bordine, amply demonstrates that this assump-tion is erroneous.
i Mr. Bordine shows that all leaks must proceed either from the spent fuel pool itself or from systems ~ con-4 nected with it.
Leaks from the pool itself, which would have to result from a puncture of the stainless steel pool liner, would. collect in drain tubes terminating in manually-operated valves.
When these valves'are opened O
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the water would be collected in the containment sump and be processed through the plant's radwaste system.
Three systems connect with the spent fuel pool:
4 the Spent Fuel Pool Cooling System (SFPCS), the demin-eralized water system and the treated waste system.
Pipe failure in the SFPCS would not drain the spent fuel pool but only the adjacent surge tank; leaked water would j
be collected in the enclosure sump.
The demineralized i
water system has one connection with the SFPCS.
In 1978 a check valve failure at this junction resulted in leakage of pool water into the demineralized water sys-3
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tem; some of this demineralized water was removed from the site in bottles.
This check valve has now been placed under lock control, periodic testing of the de-mineralized water has been instituted, and radiological clearance practices for removal of material from the site have been revised.
These safeguards preclude re-currence of the incident.
Similar leakage paths through connections from the SFPCS to the treated waste system exist; however, treated waste water is itself radioactive and is handled under procedures reflecting the require-ments of 10 CFR Parts 20 and 100.
The other two leakage incidents referred to in Mr.
O'Neill's contention involved excessive leakage from valve CV/4027 detected during routine leak rate tests.
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__. - O Any leakage of pool water from the SFPCS through this valve would go directly to the plant's radwaste system.
Mr. Bordine's testimony thus demonstrates that escaped water from all small to medium leaks from the spent fuel pool, whatever the leakage path, would either remain within containment or be processed through the plant's radwaste system.
Such leaks could not result in uncontrolled releases to the eavironment and thus pose no environmental hazards.
Mr. O'Neill has raised no-fact issue which controverts the facts established in the testilnony of Mr. Bordine; accordingly, Licensee
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is entitled to summary disposition of the contention as a matter of law.
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s STATE OF MICHIGAN
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SS.
COUNTY OF JACKSON
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UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFCRE THE ATOMIC SAFETY AND LICENSING BOARD IN THE MATTER OF
)
CONSUMERS POhER COMPANY
)
Docket No. 50-155 OLA (Big Rock Point Nuclear Plant)
) (Spent Fuel Pool Expansion)
AFFIDAVIT OF THOMAS C.
BORDINE i
My name is Thomas C.
Bordine.
My business address i
i is 1945 W.
Parnall Road, Jackson, Michigan.
I am employed by Consumers Power Company as a Staff Licensing Engineer for the Big Rock Point Plant.
I have a Bachelor of Science Degree in Mechanical Engineering from Wayne State University,
)
Letroit.
I joined Consumers Power Company in the Gas Engineer-ing Department of the Livonia, Michigan district office in June 1967 and remained at that location until December 1975 1
(excepting two years military leave in 1968-1970).
My primary responsibilities as Systems Planning Designer included directing the updating of the Gas Network Flow Analysis, as well as the formulation of proposed system networks.
In January of 1976, I transferred to the Big Rock Point Plant, Charlevoix, Michigan as Quality Assurance Engineer.
In that capacity, my responsibilities as a staff member of the site Quality Assurance Department included conducting QA Program Audits and Surveillances; performing quality reviews of
()
procedures, design and procurement documents; and monitoring and performing final close-out reviews of the site corrective action system documents.
_ 1 In June 1977, I was assigned the position of i
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Quality Assurance Superintendent, Big Rock Point Plant.
My responsibilities included organizing and directing the
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activities of the Plant Quality Assurance and Qua1ity Control staff; providi g assistance to the plant technical
~
staff to more effectively implement the QA Program; analyz-ing the effectiveness of the CA Program and recommending program improvements to QA Department Management; reporting the status of QA Program implementation at the plant to QA Department Management; and representing the QA Program and QA Department to the NRC I/E Branch personnel.
During my assignment at Big Rock, I also participated in and cct-pleted, in April 1980, an administrative certification
(])
program for Senior Reactor Operator Equivalency.
The program included extensive training in plant systems, con-trols and licensing requirements as well as reactor start-up qualification at the General Electric Boiling Water Reactor Simulator.
In August 1980, I was assigned to work at the j
Institute of NucleTr Power Operations, Atlanta, Georgia in a one-year on-1( an capacity.
My primary responsibility at the Institute was to establish a process for the development and coordination of nuclear operations management criteria.
4 Since August 1981, I have been assigned as Big Rock Point Staff Licensing Engineer at the Corporate Office in Jackson, Michigan.
I'am responsible for the coordination.and response of licensing _ activities related to the Big Rock Point Plant
}
whicn include the spent fuel expansion project.
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I am the author of the documents entitled "Testi-mony of Thomas C. Bordine Concerning O'Neill Contention II-B", " Testimony of Thomas C. Bordine Concerning O'Neill Contention II-C", and " Testimony of Thomas C. Bordine Concerning Licensing Board Question No.
1".
I believe that my educational background and work experience qualifies me to respond to tho'c contentions and question.
Where state-ments of fact are made in this testimony, they are based on my personal knowledge.
Where results of calculations performed by Consumers Power Company are discussed, I have personally reviewed and verified the calculations.
I swear that this affidavit and the testimony attached hereto are true and correct, to the best of my knowledge and belief.
% MGmL 4
Thomas C. Bordine SUBSCRIBED AND SWORN TO before me this 2nd day of October 1981.
. ( _
s/ h L Y Y (/4' [b/4 h Notay? Fublic Jackson County, Michigan My Commission expires March 26, 1983 4
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