ML20044A211
| ML20044A211 | |
| Person / Time | |
|---|---|
| Site: | Pilgrim |
| Issue date: | 06/21/1990 |
| From: | Carr K NRC COMMISSION (OCM) |
| To: | Griffin W PLYMOUTH, MA |
| Shared Package | |
| ML20044A212 | List: |
| References | |
| NUDOCS 9006280213 | |
| Download: ML20044A211 (14) | |
See also: IR 05000293/1988007
Text
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June 21
1990
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Mr. William R. Griffin
Executive Secretary
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Office of the Selectmen
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11 Lincoln Street
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Plymouth, Massachusetts
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),arr
Dear Mr. Griffin:
MRushbrook
RWessman
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l' am responding to your letter of April 24, 1990, concerning the
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' direct torus vent at the Pilgrim Nuclear Power Station.
I
referred the 12 specific questions you raised in your letter to
the Nuclear Regulatory Commission (NRC) staff, and their detailed
responses are enclosed.
Some additional background information
that may be helpful to you is also enclosed.
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I hope the information we are providing will lead to a better
understanding of the generic issues associated with venting, and,
in particular, how-they relate to the Pilgrim Nuclear Power
Station.
If you~have any further ouestions, please contact me or
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Mr. T. T.-Martin, Administrator of NRC's Region I office.
-Mr. Martin can be' reached by telephone at-(215) 337-5299.
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Sincerely,
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Kenneth M. Carr
Enclosures:
1.
Background Information
2.
Responses to Concerns
3.
4
Inspection Report No. 50-293/88-07
5.
Inspection Report No. 50-293/88-12
Originated:
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9006280213 900621
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COMMS NRCC
CORRESPONDENCE PDC
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Enclosure 1
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Backoround Information Related to Pilgrim Station's
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Direct Torus Vent System (DTV5)
On January 23, 1989, the NRC staff presented its recommendations on Mark I
containment performance improvements and other safety enhancements to the
Comission in SECY 89-017.
It represented the completion of the staff efforts
on the Containment Performance Improvement (CPI) Program for Mark I containments.
The program was established to determine what actions, if any, should be taken
to reduce the vulnerability of containments to severe-accident challenges.
From this point of view, the staff proposed that hardened vent capability would
enhance plant capabilities with regard to both severe accident prevention and
mitigation.
Some low probability scenarios in which multiple failures occur can lead to
containment failure. Containment failure from these scenarios can result in
a loss of cooling water which is used to remove decay heat.
The installation
of a hardened vent greatly reouces the likelihood of early containment failure
and, therefore, reduces the risks to the public because cooling capability is
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maintained.
For other sequences for which core melt is predicted, venting
cruld be effective in delaying containment failure and in mitigating the
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release of fission products.
Although venting of the containment is currently
included in BWR emergency operating procedures to improve the survivability of
the containment, which acts as the last barrier for an uncontrolled release of
radiation, it generally uses a vent path that includes ductwork with a low
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design pressure. Venting under high-pressure severe-accident conditions could
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fail this ductwork, release the containment atmosphere into the reactor
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building, and damage equipment or contaminate equipment needed for accident
recovery. Venting through this ductwork may hamper or complicate post-accident
recovery activities.
The installation of a reliable hardened wetwell vent
allows for controlled venting through the wetwell while providing a path with
significant scrubbing capability of fission products to the plant stack and
prevents damage to equipment needed for accident recovery. Based on the
staff's recomendation, the Comission directed the staff to allow the
licensees that elected to incorporate this plant improvement to install a
hardened wetwell vent in accordance with the Comission's regulations
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Plant specific backfit analyses were directed for the
remaining plants with Mark I containments. Where these analyses supported
imposition of a hardened vent, the staff was directed to issue orders
requiring this modification.
Prior to the Comission decision in this matter, numerous discussions with both
industry groups and individual licensees were conducted.
These discussions
included meetings with Boston Edison (the licensee for Pilgrim). The purpose
.of these discussions was to gather all available information relative to the
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hardened vent to enable the staff to make an informed decision.
During this
process, Boston Edison proposed to install the Direct Torus Vent System (DTVS).
The licensee had concluded that it had sufficient information to commit to
a specific design for hardened wetwell vents. The proposed modification was
. consistent with the staff's generic finding for Mark I plants.
Hcwever, the
' staff did not use the Pilgrim design as a test case, as is indict.ted in your
letter.
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Boston Edison described the design of the DTVS at Pilgrim in the " Report
on Pilgrim Station Safety Enhancements" of July 1, 1987, and subsequently
revised the report on August 18, 1988
The design provides a direct vent path
from the torus to the main stack, bypassing the Standby Gas Treatment System
(SBGTS). The bypass is an 8-inch line with the upstream end connected to the
pipe between the primary containment isolation valves. An 8-inch butterfly
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valve (AO-5025), which can be remotely operated from the main control room,
is added in the bypass line.
This valve is the primary containment outboard
isolation valve for the direct torus vent.
A rupture disk is also provided
downstream of this outer isolation valve.
The NRC staff conducted an inspection at the Pilgrim Nuclear Power Station on
March 2-3, 1988, and documented its evaluation in NRC Inspection Report
No. 50-293/88-07, of May 6. 1988, and Inspection Report No. 50-293/88-12,of
May 31, 1988. The s af Sn33%d the installed system and the associated
analysis acceptable.
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Enclosure 2
Response to Concerns Raised by W.R. Griffi_n,n
The following items briefly sumarize current information concerning the
hardened vent. They are organized as specific responses to issues raised in
your letter to Chairman Carr. You should note that two descriptive tems
routinely used within the industry mean the same thing:
both the " direct torus
vent" and the " hardened wetwell vent" describe the vent path to the stack.
For
purposes of the following responses, they are equivalent.
Question 1 (Q 1): What are the decontamination factors for the pool for
various isotopes?
In other words, how well does the wet
well pool scrub out the fission by-products, keeping the
radioactive particles from releasing to atmosphere?
Response:
Except for the noble gases (consisting of the isotopes of Xenon
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and Krypton), which are not retained by the pool to any
significant degree, the suppression pool is highly effective
in scrubbing out and retaining particulate and volatile fission
products.
Calculations as well as tests indicate that the sup-
pression pool would be expected to have a realistic decontamination
factor (DF) for particulate and volatile fission products of about
100, depending upon the accident sequence and the temperature of
the water. This means that about 1 percent of the particulate and
Wlatile radioactivity entering the pool would be released to the
atmosphere, and about 99 percent would be retained within the pool.
The wetwell pool is highly effective with a DF of about 100 in
scrubbing particulate and volatile fission products, but not
effective in scrubbing noble gases with a DF of 1.
Q 2:
Please provide a graph of offsite radiation doses based on the
possibility of a vacuum breaker valve remaining open at 10%, 25%,
50% and 100%.
Response:
The staff does not have the off-site radiation dose evaluation
requested in your letter. This type of failure was not considered
in the design basis for the facility since it was not considered
to be a credible event. The basis for the staff's position in
this regard is as follows.
The vacuum relief for both the drywell and wetwell is provided by
two 100 percent vacuum relief breakers located in two penetrations
in_ the wetwell containment shell. These penetrations terminate in
the reactor building, which is generally referred to as the
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Each penetration consists of a vacuum breaker and an air operated
butterfly valve in series.
During normal operation, both valves
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are closed; the vacuum breaker is maintained closed by the weight
of the disk, and the butterfly valve is maintained closed by
positive actuator air pressure.
In the event of a loss-of-coolant accident (LOCA), the increasing
wetwell pressure will add to the closing pressure of the vacuum
breaker. As a result, it is anticipated that during the entire
positive pressure history within the containment, neither valve in
the penetration will move from its closed position.
However, at
the end of the pressurization phase, there is a potential for
creating a negative pressure in containment. This would occur only
after the steam release from the reactor coolant system has ceased.
As the wetwell pressure approaches atmospheric, the butterfly
valve is openeo, thereby allowing the vacuum breaker to properly
function. The vacuum breaker would begin to open when the wetwell
pressure becomes slightly sub-atmospheric. Air from the reactor
building would restore the wetwel
pressure back to atmospheric.
The above sequence description has focused on the Design Basis
Accident (DBA). However, the sequence is equally valid for a
large number of potential severe accident scenarios.
The dif-
ferences would be limited to the pressure rise rate and the
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maximum pressure and tempertture values reached during the event.
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These differences, however, would not alter the events as described
above. Therefore for purposes of consideration of vacuum breaker
failure, the staff's conclusions can be considered applicable for
both DBA and severe accident events.
Therefore, during the entire positive pressure profile of the
event, the penetration has two closed barriers in series.
It is
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only during the end of the pressurization phase that'the
penetration is aligned into its vacuum breaker role.
Because of-
this double barrier protection and the fact that both valves are
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not expected to change position during the pressurization phase of
the event, the staff has concluded that failure of the penetration
as a leak tight barrier is not credible and need not be considered
in the design basis.
Q 3:
The NRC has recomended venting at the containment design pressure
as a minimum, or in the case of Pilgrim, at 60 psi.
Why is the
Pilgrim DTVS rupture disk set at half that, at 30 psi?
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Response:
The fact that the Pilgrim DTVS rupture disk is designed to rupture
at 30 psi is not related to the NRC's recommendation that specified
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the venting pressure at the containment design pressure. The set
pressure for the rupture disk does not control the venting pressure
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because there are two closed isolation valves in the flow path.
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These two valves are normally closed and will be cpened manually
by the operator if venting is needed.
Pilgrim's venting pressure
in this case is consistent with the recommendations contained in
Ecergency Procedure Guidelines (EPG), Revision 4
These guidelines
have been approved by the staff. The maximum containment pressura
at which the operators are expected to open the vent valve is 56 psig
(not 60 psi), which is consistent with the NRC recommendation on
venting pressure.
The rupture disk is designed to serve as an additional leakage
barrier at pcc:svrat below 30 psi.
It is designed to open
below the containment design pressure, but will be intact up to a
pressure equal to or greater than those pressures that cause an
automatic containment isolation during any accident conditions,
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Therefore, its presence in the line can effectively eliminate the
negative consequences of inadvertent actuation of the vent valves
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at pressures below 30 psi. The set pressure of 30 psi for the
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rupture disk satisfies these design objectives.
04:
What is the minimum containment pressure allo W by procedures
at which the operators could open the DTVS outbotrd containment
valve, A0-5025?
Ratnanse:
Use of the direct torus vent will be in accordance with approved
EPG requirements and will be controlled by Emergency Operating
Procedures (EOPs). There is no specified minimum containment
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pressure allowed by the BWR Owners Group EPGs, Revision 4 at
which the operators could open the DTVS outboard. containment valve.
There is a primary containment pressure limit (PCPL) of 56 psig.
Plant-specific supporting analyses are used to indicate when the
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operators should begin the venting procedure. These analyses
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. considered a number of plant parameters, including the pressure
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rise rate. These actions ensure that venting is used only if
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needed, that the conditions are beyoi.d the des'gn-basis-accident
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assumptions, and that the pressures in the m tsinment do not
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exceed the PCPL limit.
Q 5:
Please provide information on the reliability of thi. hydrogen and
oxygen concentration monitors at Pilgrim.
What percentage of the
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time have both systems been accurately functioning?
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Response:
The post-accident hydrogen / oxygen analyzers were installed in
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January 1985 as part of the post-TMI design modifications.
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Since the installation, one train (of two) was inoperable for
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two days in November 1985, and one train was inoperable for four
days in January 1986, for a total of six days. At no time
were both trains inoperable simultaneously. Technical Specification 3.7.A.7.c allows the reactor to operate for up to 7 days if one
train is inoperable.
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In addition, the containment atmospheric oxygen analyzer, which
monitors the oxygen concentration during normal operation, has
been extremely reliable. The plant staff conservatively estimated
this analyzer to have a reliability that exceeds 98 percent.
Q 6:-
Does the NRC concur that the use of the DTVS does not involve an
unreviewed safety question?
Response:
Yes. As documented in NRC Inspection Report No. 50-293/88-07,
dated May 6,1988, the NRC inspected the installed DTVS design
configuration and the licensee's evaluation and determined that
they were acceptable.
Venting has been approved under previous
versions of the EPGs. The direct torus vent is initiated by
procedures under conditions specified by the EPGs. Because the
outboard valve, A0-5025, is sealed closed and subject to leak
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testing, this valve satisfies the provisions of 10 CFR Part 50,
Appendices A and J, which are the regulations for containment
isolation and leak testing, respectively.
Therefore, the NRC
concurred that the use of the DTVS does not involve an unreviewed
safety question.
-Q 7:
Does the NRC concur that the use of the DTVS does not require
changes to Pilgrim's Technical Specifications?
Response:
Yes', the NRC agrees that the use of the DTVS does not require
changes to Pilgrim's Technical Specifications. Our inspection
reports, which were noted in the previous responses, incluced
consideration of possible TS changes, and we determined that none
were needed.
0 8:
Does the NRC judge the DTVS to improve the safety at Pilgrim?
Response:
Yes.
The DTVS provides an improv1d containment venting capability-
for decay heat removal. The DTVS will prevent the majority of
postulated loss of decay heat removal sequences from resulting
in core melt and will mitigate the consequences of the residual
sequences involving core melt where venting through the suppression
pool is found necessary. Additional safety benefits of DTVS are
discussed in the previous background paragraphs.
Q 9:
Does the NRC conclude that the installation and use of the DTVS
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are acceptable under the provisions of 10 CFR 50.597
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Response:
Yes. As we noted in the response to Question 6, the staff inspected
the design of DTVS at Pilgrim and found the installed system and
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the associated anal / sis acceptable.
Venting had been approved
under previous versions of the EPGs. The direct torus vent is
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initiated by procedures under conditions specified by the EPGs.- In
addition, the installation or us6 of the direct torus vent will not
increase the probability of a new accident. Therefore, the
installation and use of the DTVS are acceptable under the
provisions of 10 CFR 50.59.
Furthermore, in a supplemental assessment of October 12, 1988, the
NRC staff conclJded that the Safety Enhancement Program (SEP)
modifications being implemented in accordance with 10 CFR 50.59,
including the tTV modification, would enhance the overall plant
safety and perf ormance of Pilgrim.
Q 10:
Does the NRC conclude that Boston Edison has adequately considered
the technical issues germane to the DTVS?
Response:
Yes. Based on the noted inspections and reviews of the Pilgrim SEP,
the NRC staff concludes that the safety issues associated with
the DTVS have been adequately considered.
Q_11:
Why was.the automatic reclosure on high radiation of valve
A0-5025 deleted during the design revision of the system?
Response:
The reclosure of valve A0-5025 was deleted because this reclosure,
if performed at high radiation levels, would isolate the vent flow
path when venting is needed to mitigate the overpressure challenge.
Thus, automatic reclosure could defeat the purpose of the vent
design.
Q 12:
Generic Letter 89-16 indicates some benefits of a hardened wet
well vent to reduce core damage frequencies during SB0 [ station
blackout) and ATWS [ anticipated transient without scram] accident
scenarios.
Is this true for Pilgrim?
Response:
Yes. The isolation valves, A0 5025 and A0-5042B, are designed
with ac independent power supplies. These two v.alves are powered
from essential de powsr and are backed up with diverse nitrogen
actuation capability:
Therefore, in case of an SB0 event, the
valves would be available for venting. The venting concept is
mainly designed to slow overpressure transients of the contain-
ment. During some ATWS events, the pressure in the containment
will rapidly increase. Venting pressure could be reached in a
matter of minutes rather than hours. Therefore, venting may not
prevent containment failure because of the high containment
pressurization rate but would provide additional time to scram the
reactor and delay the core melt.
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Mr. William R. Griffin
Executive Secretary
Town of Plymouth
Office of the Selectmen
11 Lincoln Street
Plymout , Massachusetts 02360
Dear Mr.
iffin:
.
I am respond 1pg to your letter of April 24, 1990, concerning the direct
torus vent at'the Pilgrim Nuclear Power Station.
You raised 12 questions
concerning a vehiety of issues regarding the direct torus vent. The issues
decontamination factors of the sup)ression pool dose
arelistedasfollows:
consequences from the cpening of a vacuum breaker valve, tle pressure of the
rupture disk, the minimuni containment pressure allowed by procedures at which
the operators could 6 pen ths vent valve, the reliability of the hydrogen and
oxygen monitors at Pil' grim, anreviewed safety questions, changes to the
saf sty improvement, compliance of 10 CFR 50.59,
Technical Specifications,5 corsideration, automatic reclosure of the vent
adequacy of the licensee
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valve, and the benefits du' ring the accident scenarios of station blackout and
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anticipated transient withou scram. The staff has responded to each of your
questions in the enclosure.
Sincerely,
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K'enneth M. Carr
Enclosure:
As stated
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This correspondence addresses policy issues previously resolved by the
Commission, transmits factual information, or restates Commission policy.
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Mr. William R. Griffin
Executive Secretary
Town of Plymouth
Office of the Selectmen
11 Lincoln Street
.
Plymouth,
ssachusetts 02360
Dear Mr. Grif
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I am responding to your letter of April 24, 1990, concerning the direct
torus vent at theVilgrim Nucicar Power Station. You reited twelve questions
conctrning a variety,of issues regarding the direct torus vent. The issues
are listed as follows;
decontamination factors of the sup'ression pool, dose
a
consequences from the %pening of a vacuum bruker valvo, t1e pressure of the
rupture disk, the minimog containment pressure allcwed by procedures at which
the operatort could open the vent valve, the reliability of the hydrogen and
oxygenrnonitorsatPilgrimhunreviewedsafetyquestions,changestothe
Technical Specifications,s consideration, automatic reclosure of the vent
.
safety improvement, cornpliance of 10 CFR 50.59,
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adequacy of the licensee
valve, and the benefits during\\the accident scenarios of station blackout and
anticipated transient without scVam. The staff has responded to each of your
questions in the enclosure.
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S,incerely,
\\
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Kennet
M. Carr
Enclosure:
As stated
This correspondence addresses puli
issues previously resolved by the
Commission, transmits factual infonc.cion, or rehtates Commission policy.
DISTRIBUTIO_N
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Docket File 50-293
JBlaha
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Mr. William R. Griffin
Executive Secretary
Office of the Selectmen
Town of Plymouth
11 Lincoln Street
Plymouth, Massachusetts 02350
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Dear Mr. Griffin:
,
I am respondin
o your letter of April 24, 1990 concerning the direct
torus vent at the Pilg
Nuclear Power Station. The NRC staff has responded
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to your questions in the
closure to this letter.
Sincerely,
Kenneth M. Carr
Enclosure:
As stated
This correspondence addresses policy issues previously resolvecf by the
Consnission, transmits factual information, or restates Consnission policy.
'
DISTRIBUTION
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Mr. William R. Griffin
Executive Secretary
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Office of the Selectmen
Town of Plymouth
11 Lincoln Street
Plymouth, Massachusetts 02360
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Dear Mr. Griffir)\\
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I am respondin'g to your letter of April N,1990 concerning the direct
torus vent at the Pilgr\\ m Nuclear Power Station. The NRC staff has responded
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to your questions in the enclosure to this letter.
Sincerely,
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Chairman Carr
Enclosure:
As stated
This correspondence cd6resses policy 1_ssues previously resolved by the
Commission, transmits factual information, or restates Comission policy.
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EDO Principal Correspondence Control
FROM:
DUE: 05/08/90
EDO CONTROL: 0005403
DOC DT: 04/24/90
FINAL REPLY:
William R. Griffin
? Town.of-Plymouth
TO:
Chairman Carr
.FOR SIGNATURE OF:
CRC NO: 90-0440
Chairman Carr
.DESC:
ROUTING:
- QUESTIONS CONCERNING THE DIRECT TORUS VENT NOW
Taylor
. OPERATIONAL AT THE PILORIM NUCLEAR POWER PLANT
Sniezek
'
Thompson
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DATE:-04/27/90
Blaha
.
Beckjord, RES
LASSIGNED TO:
CONTACT:
Scinto. 000
NRR.
Murley
'
SPECIAL INSTRUCTIONS OR REMARKS:
'NRR' RECEIVED:
APRIL 27, 1990
ACTION:
! DST;'gHADAN1?
NRR ROUTING:
MURLEY/MIRAGLIA
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OFFICE OF THE SECRETARY
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CORRESPONDENCE CONTROL TICKET
- - -
PAPER NUMBER:
CRC-90-0440
IDGGING DATE: Apr 26 90
-ACTION OTFICE:
AUTHOR:
William R. Griffin
AFFILIATION:
MA (MASSACHUSETTS)
LETTER DATE:
Apr 24 90
FILE CODE: ID&R-5 Pilgrim
!
SUIL7ECT:
Questions concerning the direct torus vent now
operational at the Pilgrim nuclear power plant
'
ACTION:
Signature of Chairman
-
DISTRIBUTION:
RF, DSB, Chairman
.
SPECIAL HANDLING: None
NOTES:
,
DATE DUE:
May 10 90
SIGNATURE:
DATE SIGNED:
.
AFFILIATION:
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