ML20236T227
| ML20236T227 | |
| Person / Time | |
|---|---|
| Site: | Limerick  |
| Issue date: | 11/23/1987 |
| From: | Doering J, Wiley J PECO ENERGY CO., (FORMERLY PHILADELPHIA ELECTRIC |
| To: | |
| Shared Package | |
| ML20236T163 | List: |
| References | |
| GL-85-19, OLA, NUDOCS 8712010086 | |
| Download: ML20236T227 (64) | |
Text
-
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o-UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION Before the Atomic Safety and Licensing Board In the Matter of
)
)
Philadelphia Electric Company
)
Docket No. 50-352-OLA
)
(TS Iodine)
(Limerick Generating Station,
)
Unit 1)
')
AFFIDAVIT OF JOHN DOERING AND JOHN S. WILEY IN SUPPORT OF LICENSEE'S MOTION FOR
SUMMARY
DISPOSITION Messrs. Doering and Wiley being duly sworn according to law come forth and say:
1.
My name is John Doering.
I am employed by the Philadelphia Electric Company
(" Philadelphia Electric Company" or " Company").
I am presently assigned to the Limerick Generating Station where I
am responsible for management and oversight of plant operations, engineering and chemistry support.
I hold a Bachelor of Science degree from the University of Pennsylvania.
2.
I have previously served in the capacity of Engi-neer-Operations, Limerick Generating Station and Engi-neer-Plant Test, Peach Bottom Atomic Power Station.
I presently hold Nuclear Regulatory Commission Senior Reactor Operator license No. 10144.
I have held a senior reactor operator license for 8 years.
3.
Part of the responsibilities in my present position include the review of matters for deportability to the p[
bR OC Q
l 3
.- 2 --
Nuclear Regulatory Commission and' the-review of proposed Technical Specification amendments.
I serve on-the Station Operation Review Committee.
As a result, I.am familiar with the August 19, 1986 Application for Amendment of Facility Operating License NPF-39 relating to the reporting of iodine spikes and licensing requirements in general.
I am familiar 1
with plant operation and chemistry requirements.
A copy.of the statement of my professional qualifications, Attachment 1 hereto, is incorporated by reference herein.
4.
My name is _ John Wiley.
I am ' Director of the Nuclear Plant Chemistry Section for the Philadelphia Elec--
tric Company.
In that position, I am responsible for the technical direction of chemistry programs at.the Company's nuclear facilities.
Prior to that' time, I was assigned to the Limerick ' Generating Station as Senior Chemist where I
)
was in charge of the Chemistry Department, heading a staff of approximately 35 chemists and technicians.
I hold Bachelor of Science, Master of Science and Ph.D degrees in Chemical Engineering from Iowa State University.
I am also a graduate of the Oak Ridge School of Reactor Technology.
5.
My previous work experience included assignments with the General Electric Company related to boiling water reactor fuel development and' chemistry.
I am knowledgeable in plant chemistry requirements and the phenomenon of iodine spiking and am familiar with the pending Technical Specification change request related to that subject.
Q.
6.
I am a member of the American Nuclear - Society, American Society of Mechanical Engineers and a ' Registered
~
Professional Engineer in Pennsylvania.
A copy-of the statement of my professional qualifications, Attachment 2 hereto, is incorporated by reference herein.
7.
We have'been asked.by Philadelphia Electric Company to respond to the contention stated by the Atomic Safety and Licensing Board in its October 9, 1987 Memorandum and Order 1
as follows:
l Consolidated Contention.
-The.-proposed amendment to the Licensee's technical specifications would~ downgrade reporting requirements-for iodine spikes which would have an adverse effect on public health and safety.
Bases.
The change in the. reporting requirements would eliminate or decrease Special Reports and Licensee Event Reports on. iodine
- spiking, and thus would decrease the regulatory control exercised by the NRC, would permit a i
situation where Licensee could ' release i
radioactive iodine-in excess of the one-time release
- limits, and, in not requiring the reporting of such-re-leases, except on an annual basis, would~
endanger the health and safety of the uninformed public.
8.
We have carefully examined and analyzed the con-tention of the Air and Water Pollution Patrol and Mr.
Anthony.
Our consideration of the contention utilizes our i
extensive experience related to the design, operation, and licensing requirements of nuclear power plants.
.As dis-cussed in more detail below, our conclusion is that the proposed amendment would not change reporting requirements 1
related to iodine spikes such that there would be an adverse effect on public health and safety.
We base this conclusion on the improvements in fuel design and management which have l
resulted in an extremely low probability of iodine spiking resulting from the operation of the Limerick Generating Station.
- Moreover, existing reporting requirements con-l tained in the NRC's Regulations which are entirely unaffect-l ed by the proposed amendment would provide sufficient j
l information to the NRC to enable it to independently analyze and evaluate any iodine spikes.
Moreover, there are early notification requirements which would alert the NRC, State and local officials to iodine levels in excess of Technical Specification permitted values.
The proposed amendment i
1 leaves unchanged the release limits and reporting require-l ments for radioactive effluents from the site and does not affect accident calculations associated with iodine levels in the reactor coolant.
Thus, as discussed below in further detail, we conclude that the contention has no merit.
l 9.
On September 27,
- 1985, the Nuclear Regulatory Commission
(" Commission" or "NRC")
Staff issued Generic Letter 85-19, regarding the requirements for reporting l
iodine spikes during normal plant operation.
See Attachment 1
I 3.
That letter requested licensees to file a request for amendment to their operating licenses to incorporate the NRC model Technical Specifications relating to iodine spikes.
On August 19, 1986, in response to that request, Licensee filed an application with the NRC requesting, changes to the
/
Technical Specifications contained in Appendix A of Facility Operating License NPF-39 for Limerick Generating Station, Unit 1.
See Attachment 4.
A copy of the present Technical Specifications which govern the operation of the Station related to-iodine spiking is attached hereto as Attachment 5.
10.
The application would amend Technical Specification 3/4.4.5 to-eliminate the requirement to shut down the facility if the coolant iodine activity limits were exceeded for 800 hours0.00926 days <br />0.222 hours <br />0.00132 weeks <br />3.044e-4 months <br /> in any 12-month period.
The application also requested that the requirements to submit special 30- and 90-day reports if the Technical Specification iodine coolant 1
activity of 0.2 microcurie per gram were exceeded be changed to require that such information be included'in the annual report required by the Technical Specifications.
Section 6.9.1.5(d) was to be added to implement this requirement.
In addition, the requirement to submit a special report within 30 days if the iodine coolant activity were. exceeded for a period of 500 hours0.00579 days <br />0.139 hours <br />8.267196e-4 weeks <br />1.9025e-4 months <br /> in any consecutive 6-month period was proposed to be eliminated as unnecessary.
11.
The proposed amendment does not change the require-ments for specific activity limits for iodine in the reactor coolant.
As is evident from a comparison of Attachments 4 and 5,
both require that the reactor be placed in a hot shutdown condition with the main steamline isolation valves closed within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> if the specific activity of the l
I primary coolant exceeds 4
microcuries per gram dose l
-_ a
.1 l
equivalent I-131 (hereinafter "microcuries per gram").
Both the original and revised Technical Specification allow operation for only 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> if the specific activity in the primary coolant is greater than 0.2 microcurie per gram but less than or equal to 4 microcurie per gram.
If within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> the iodine activity level cannot be reduced to 0.2 microcurie per gram or less, the unit must be placed in a hot shutdown condition with the main steamline isolation valves closed within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> in either case.
The requested I
changes involve no modification to the Station radioactive effluent release limits or in reporting requirements related to plant radioactive effluents.
i 12.
It is extremely unlikely that the triggering level for the Technical Specification in question, e.g.,
0.2 microcurie per gram, would ever be reached during the life of the unit.
An iodine spike is defined as an increase and.
subsequent decrease in iodine dose equivalent following a change in reactor power.
The magnitude and duration of the i
spike is a function of the power transient and clad integri-ty.
High levels of iodine in the reactor coolant were encountered by boiling water reactors operating in the early 1970's.
These high levels resulted from fuel cladding failures caused by moisture trapped inside the fuel rod, pellet-clad interactions and crud-induced corrosion.
Improvements in the design of the nuclear fuel such as internal moisture
- getters, barrier
- fuel, improved clad manufacturing techniques, improved fuel management practices, and the replacement of the older fuel assemblies gradually eliminated the failed fuel and the resulting higher levels of iodine in operating reactors.
13.
We have confidence that the 0.2 microcurie per gram threshold value would not be reached at Limerick.
This confidence has been gained as a result of our review of the experience to date monitoring iodine levels at other boiling water reactors and at Limerick.
Since startup, Limerick has
-5 averaged only 8 x 10 microcurie per gram of iodine in the coolant for the first cycle of operation during Operational Condition 1 (') 75% power) with a maximum value of only 1.2 x
-4 10 microcurie per gram during this period.
The average for the first fuel cycle is only 0.04% of the threshold value of 0.2 microcurie per gram contained in the Technical Specification.
The average level of iodine in the reactor coolant for the second cycle through November 13, 1987 is 5
-5 x 10 microcurie per gram.
The industry 1986 median value for iodine coolant activity in boiling water reactors was
-3 1.5 x 10 microcurie per gram which is still only 0.75% of i
the threshold value of 0.2 microcurie per gram.
14.
Sampling for iodine coolant activity is conducted at the Station in accordance with Technical Specification 4.4.5.
During operation, the frequency of iodine sampling is daily.
The level of iodine coolant activity is reviewed periodically by the plant chemistry staff for trends and accuracy.
The Station has established an administrative limit of 0.002 microcurie per gram which is 1% of the
8-Technical Specification limit.
Were this administrative limit exceeded, this information would be discussed at the daily chemistry meeting held at the Station, management notified and available courses of action considered.
As the Director, Nuclear Plant Chemistry, I review the data monthly for trends, and trends are sent to department management on a
monthly basis.
The information is also transmitted periodically to the fuel supplier for evaluation.
This data also is transmitted to INPO for inclusion in its "Perfor-mance Indicator Report."
The Company also participates in an industry fuel surveillance program which was established to detect fuel performance anomolies and to confirm expected performance.
15.
There are a
number of reporting requirements contained in the NRC regulations which would require the Station to give early notice to the NRC and then provide complete information regarding any high iodine levels encountered in the coolant.
In addition to these reports, the NRC has assigned Resident Inspectors to review operation of Limerick Unit 1.
Periodic inspection reports which highlight the Station operations, including any adverse trends, are routinely prepared by the Resident Inspectors and forwarded to the Region-and Headquarters for analysis and evaluation.
NRC Inspection specialists also routinely inspect the facility and review among other things, plant chemistry operations.
These inspections are similarly documented, reviewed and are also made public.
.. 16.
The first reporting requirement relevant to iodine coolant levels is 10 C.F. R.
S50. 73 (a) (2) (i) which requires that a Licensee Event Report be filed following completion of any plant shutdown required by the~ Technical Specifica-tions.
Under the proposed (as well as the present) Techni-cal Specifications, should the iodine coolant activity exceed 4 microcurie per gram or 0.2 microcurie per gram for 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, a plant shutdown would be required and a Licensee Event Report submitted to the NRC.
In accordance with S50.73(b), the LER must include the details surrounding the event, its cause and corrective accion and provide a refer-ence to previous similar events.
LERs, being public docu-ments, are placed in the Public Document Room in Washington, l
D.C.
and the Local Public Document Room in Pottstown, Pennsylvania near the facility.
17.
Second, any plant shutdown required by'the Techni-cal Specifications would require one hour notification of the NRC pursuant to 550.72 (b) (1) (i).
Such notification l
would be from the Limerick Control Room to the NRC Op-erations Center and would take place via dedicated telephone line.
The NRC Operations Center would then be responsible for directing NRC response and ensuring that NRC management was informed.
18.
The third means by which the NRC and the public l
would be notified of iodine levels exceeding the level of 0.2 microcurie per gram is via the initiation of the Station l
The Company has designated 0.2 microcurie l
1
_ 10 -
j per gram ~1evel as an action level in accordance with 10 i
C.F.R.
S50.47 (b) (4) and NUREG-0654, Rev.
1, Criteria for Preparation and Evaluation of Radiological Emergency Re-l sponse Plans and Preparedness in Support of Nuclear Power d
Plants, Appendix 1,
page 1-5, Item 3.b.
Reaching such a level would require the declaration of an Unusual Event and implementation of the Station Emergency Plan (see Attachment 7).
In accordance with the ' Plan and implementing proce-dures, State and local officials would be notified within 15 l
minutes and the NRC notified immediately thereafter.
'See 10 C.F.R. Part 50, Appendix E Section D.3 and 50.72 (a) (1) (i).
19.
Under the requested Technical Specification, there l
are a number of. overlapping mechanisms in place to assure that information related to fuel failures is available to the NRC and made known to the public.
Elimination of the special reports related to iodine coolant levels would not impede the flow of information to the NRC.
For ' example,
inasmuch as the Technical' Specifications permit' operation at iodine coolant levels between 0.2 and 4.0 microcurie per gram for at the most 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> before a plant shutdown is mandated, it would require a minimum of 11 such episodes prior to reaching the 500 hour0.00579 days <br />0.139 hours <br />8.267196e-4 weeks <br />1.9025e-4 months <br /> mark, which under the present Technical Specifications would require the submittal of a special report.
However, such special report would not need to be immediately filed, but could be submitted up to 30 days after the 500 hour0.00579 days <br />0.139 hours <br />8.267196e-4 weeks <br />1.9025e-4 months <br /> mark were reached.
Were the level of iodine coolant to exceed 0.2 microcurie per gram, it
i
_ 11o would require the - declaration L of o an; Unusual: Event. : Thus, f
s there~would be at'least'll declarations of-an--Unusual Event
.j in accordance' with the Station. Emergency Plan prior to
.l reaching the ~ 500 hour0.00579 days <br />0.139 hours <br />8.267196e-4 weeks <br />1.9025e-4 months <br /> mark.
Thus, the' requirement for la special report is extraneous and unnecessa'ry.
The. Licensee and NRC would _ have. focused upon this matter long.before 1 the special report were due..'
Similarly, the reporting requir~e-ments contained in the NRC Regulations woul'd more than compensate for the elimination of the short-term special-reports.
20.
We have also examined.the bases'for the contention s
as stated by the interveners.-
These bases.contain.certain erroneous statements and there is nothingl in them which would support the contention.
Initially, contrary to ; the I
statement in the bases, for Limerick, there are no. Licensee-l i
Event Reports which are being sought to be eliminated by the.
1 proposed amendment.
21.
The bases for the consolidated contention allege that. a situation could be permitted under. the requested Technical Specifications where the licensee could release l
radioactive iodine in excess of the one time release limits for the Station.-
- However, reports' related ~to offsite releases and the release limits themselves are governed by other Technical Specifications and the NRC regulations which l
are totally unaffected by the requested-changes.
Moreover,-
j i
the releases from the facility are governed by 10 C.F.R.
Part 20, 10 C.F.R.
550.34a and 10 C.F.R. Part 50, Appendix I; 1
l i'n
' a 1
which could not be -increased by the requested Technical
]
Specification change.
22.
In any event, in accordance-with the Station Emergency Plan, any radiological release above regulatory limits would also require' initiation of the' Plan'and immedi-j ate notifi ation of the NRC and State and local authorities.
(See Attachment'6).
23.
If the bases are addressing the consequences of design basis accidents which may be limited by the iodine level in the reactor coolant, this has no relevance to the contention.
The NRC Safety Evaluation in support' of the issuance of an operating license, NUREG-0991, the relevant 1
i L
pages of which are attached (Attachment 7), considers ar.a I
design basis accident the-failure of a main steamline.
The
.1 calculated thyroid dose resulting from this accident is controlled by the level of iodine in the coolant.
This calculation uses as input assumptions the 0.2 and 4.0 microcurie per gram levels of iodine in the reactor coolant l
contained in the Technical Specifications.
These levels which are below 10 C.F.R.
Part 100 guideline values are totally unchanged by the requested amendment.
Therefore, the calculated accident doses would. be unchanged by the grant of the amendment.
24.
For the reasons stated above, the contention has no merit and the bases are without foundation.
1
o
. /
/ John Dtsliiring
)
w Subscribed and sworn to before'me this &3 day of November, 1987 fkw Y%., J Notary Public g#
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f.1.a.;,,i. N:.:.ut.:r.J t.;a ;ist e. $f !!;: aries l
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i John Wileg
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Subscribed and sworn to before me this 20" day of November, 1987 I
\\ s]e AW ) C A<amR'
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, Notary P'ublic is p y.FRANKUM
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.MyS&10M ORIGINAL WILL BE FORWARDED TO NRC WHEN RECEIVED.
1 i
i
1 i
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ATTACHMENT 1 PROFESSIONAL QUALIFICATIONS JOHN DOERING PHILADELPHIA ELECTRIC COMPANY j
My name is John Doering.
I am employed by Philadelphia Electric Company.
I am presently assigned to the Limerick Generating Station where I am responsible for management and oversight of plant operations, engineering and chemistry l
1 support.
I have a Bachelor of Science degree in Mechanical Engineering from the University of Pennsylvania.
I present-j ly hold a Nuclear Regulatory Commission Senior Reactor operator License and have held such a license for eight years at Limerick and the Peach Bottom Atomic Power Station.
l I have been employed by Philadelphia Electric Company since 1972 snd have been involved in nuclear plant op-erations since 1973, accumulating 15 years of boiling water l
reactor operating experience.
I have served on the Boiling Water Reactor Owners Group Emergency Procedures Committee and have played a major role in the development of the Generic Boiling Water Reactor Symptomatic Emergency Proce-dures Guidelines.
l
ATTACHMENT'2 PROFESSIONAL QUALIFICATIONS JOHN-S..WILEY-PHILADELPHIA ELECTRIC COMPANY l
My name-is John S.
Wiley.
.I am presently Director, l
Nuclear Plant Chemistry Section. for the Philadelphia Elec-tric Company.
In that capacity, I.am responsible for '^.he technical direction of chemistry programs at the Company's nuclear power stations.
I hold B.S.,
M.S.
and Ph.D. degrees in chemical engi-neering from Iowa State University.
I am also a graduate of the Oak Ridge School of Reactor Technology.
I have been employed by Philadelphia Electric Company since 1983..
Prior to< assuming my present position, I was senior chemist at the Limerick Generating Station in charge of the Chemistry Department.
I managed chemistry operations until approximately eight months after the unit reached-commercial operation, and I headed a staff of.35 chemists and technicians.
I have a total of approximately 28 years of experience relating to power plant technology, and chemistry, including radiation chemistry.
I was employed by the General Electric Company for 17 years with responsibility for 'various projects related to power plants including BWR fuel.
2 s
development and radiation chemistry in boiling water reactors.
I am a member of the American Nuclear Society, American Society of Mechanical Engineers and am a
Professional Engineer in the Commonwealth of Pennsylvania,
)
l l
I k
i j
i ATTACHMENT 3
)
[ s ucy,%
UNITED STATES
- ,,.,}
NpQLEA$ REGULATORY COMMISSION y,
l 1psHWGTON. D. C. 20545 O \\,./
Seotember 27, 1985 a
TO ALL LICENSEES AND-MPLICANTS FOR OPERATING POWER REACTORS ANS!
0F CONSTRUCTION PERMITS FOR POWER REACTORS Gentlemen:
SUBJECT:
Reporting Requirements on Primary Coolant Iodine Spikes (Generic Letter No. 85-19 )
Generic Letter No. 83-43 was issued on December 19, 1983, to provide guidance i
on Technical Specification revisions required as the result of the revisions I
to 10 CFR 50.7? (Imediate Notification Requirements of Significant Events at Operating Nuclear Power Reactors) and of implementation of 10 CFR 50.73 (Licensee Event Report System). That generic letter discussed changing the
)
requirement from a Licensee Event Report to a Special Report for operating conditions where the specific activity limits of the reactor coolant are i
exceeded.
As part of our continuing program to delete unnecessary reporting require-ments, we have reviewed the reporting requirements related to primary coolant specific activity levels, specifically primary coolant iodine spikes. We have O
determined that the reporting requirements for iodine spiking can be J
reduced from a short-tem report (Special Report or Licensee Event Report) to an item which is to be included in the Annual Report. The information to be included in the Annual Report is similar to that previously required in the Licensee Event Report but has been changed to more clearly designate the results to be included from the specific activity analysis and to delete the information regarding fuel burnup by core region.
In our effort to eliminate unnecessary Technical Specification requirements, we have also detemined that the existing requirements to shut down a plant if coolant iodine activity limits are exceeded for 800 hours0.00926 days <br />0.222 hours <br />0.00132 weeks <br />3.044e-4 months <br /> in a 12-month period can be eliminated. The quality of nuclear fuel has been greatly l
improved over the past decade with the result that nomal coolant iodine activity (i.e. in the absence of iodine spiking ) is well below the limit.
1 Appropriate actions would be initiated long before accumulating 800 hours0.00926 days <br />0.222 hours <br />0.00132 weeks <br />3.044e-4 months <br /> above the iodine activity limit.
In addition, 10 CFR 50.72(b)(1)(ii) reouires the NPC to be immediately notified of fuel cladding failures that exceed expected velues or that are. caused by unexpected factors.
Therefore, this Technical Specification limit is no longer considered necessary on the basis that proper fuel management by licensees and existing reporting requirements should preclude ever approaching the limit.
Licensees are expected to continue to monitor iodine activity in the primary coolant and take responsible actions to maintain it at a reasonably low level fi.e., accumulated time with high iodine activity should not approach O;-
800 hours0.00926 days <br />0.222 hours <br />0.00132 weeks <br />3.044e-4 months <br />).
1 l
l
Enclosed are model Technical Specifications in Standard Technical Specification (STS) fomat shoving the revisions that may be used in a submittal of proposed Technical Specifications or proposed changes to existing Technical Specifications.
These changes will also be incorporated in the next revision of the STS for all nuclear power reactor vendors. The changes are indicated by a line in the A
margin of the Action Statement for the Limiting Condition for Operation.
Technical Specification amendment request should be submitted to the NRC for each facility which currently has Technical Specification reporting requirements upon exceeding coolant iodine activity limits or which has a requirement to shut down after 800 hours0.00926 days <br />0.222 hours <br />0.00132 weeks <br />3.044e-4 months <br /> with iodine above the limit. Your request should include appropriate changes to the bases section of your Technical Specifi-cations.
As a matter of information, when Technical Specifict. tion changes or other prrposed license amendments and approvals N.e., proposed facility modifications requiring i
MP.C approval) are required as a result of this or crother generic letter, they are subject to the fee provisions of 10 CFR 170 and a $150 application fee should 1
accompany ycur request (see 10 CFR 170.12(c) and 170.21).
If you have any questions relating to this subject, pleese contact M. Virgilio of my staff on (301 a92-8947).
l This request has been approved by C W Clearance Number 3150-0011, which expires September 30, 1936.
Sincerely, Huth
. Thompson.
- tor,
(
D for of Licensing
Enclosure:
Vedel Technical Specifications showing Fevisions to STS Reporting Requirements for Primary Coolant Specific Activity" List of Generic Letters l
.i i
+
g 344.4.8 SPECIFIC ACTIVITY j
LIMITING CONDITION FOR OPERATION l
3.4.8 The specific activity of-the reactor coolant shall be limited to:
i a.
Less than or equal to 1 microcurie per gram DOSE EQUIVALENT I-131, I
and j
b.
Less than or equal to 100/l micr'oCuries per gram of gross radioactivity.
APPLICABILITY:
MODES 1, 2, 3, 4, and 5.
ACTION:
MODES 1, 2 and 3*:
a.
With the specific activity of the reactor coolant greater than 1 microcurie per gram DOSE EQUIVALENT I-131 for more than 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> during one continuous time interval or exceeding the limit line shown on Figure 3.4-1, be in at least HOT STANDBY with T**8 less than 500*F within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />; and b.
With the specific activity of the reactor coolant greater than 100/l microcuries per gram, be in at least HOT STANDBY with T"V9 less than 500*F within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
MODES 1, 2, 3, 4, and 5:
With the specific activity of the reactor coolant greater than i
1 microcurie per gram DOSE EQUIVALENT I-131 or greater than 100/l micro-Curies per gram, perform the sampling and analysis requirements of Item 4.a) l' of Table 4.4-4 until the specific activity of the reactor coolant is restored to within its limits.
SURVEILLANCE REQUIREMENTS 4.4.8 The specific activity of the reactor coolant shall be determined to be i
within the limits by performance of.the sampling and analysis program of Table 4.4-4.
"With T,yg greater than or equal to 500*F.
W-STS 3/4 4-28
=__ _ _ a
l REACTOR COOLANT SYSTEM 3/4.4.8 SPECIFIC ACTIVITY i
LIMITINGCONiilTIONFOROPERATION s
0 i
3.4.8 The specific activity of the primary coolant shall be limited to:
a.
Less than or equal to 1.0 microcurie / gram DOSE EQUIVALENT I-131', and
~
b.
Less than or equal to 100/I microcuries/ gram.
]
APPLICABILITY:
MODES 1, 2, 3, 4 and 5.
1 ACTION:
MODES 1, 2 and 3*:
a.
With the specific activity of the primary coolant greater than l
1.0 microcurie / gram DOSE EQUIVALENT I-131 for more than 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> during one continuous time interval or exceeding the limit line shown on Figure 3.4-1, be in at least HOT STANDBY with T less i
avg than 500*F within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
j b.
With the specific activity of the primary coolant greater than 100/I microcuries/ gram, be in at least HOT STANDBY with T**U less than 500*F within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
MODES 1, 2, 3, 4 and 5:
With the specific activity of the primary coolant greater than 1.0 microcurie /
gram DOSE EQUIVALENT I-131 or greater than 100/l microcuries/ gram, perform the i
sampling and analysis requirements of item 4 a)' of Table 4.4-4 until the speci-fic activity of the primary coolant is restored to within its limits.
)
SURVEILLANCE REQUIREMENTS 4.4.8 The specific activity of the primary coolant shall be' determined to' be within the limits by performance of the sampling and analysis program of Table 4.4-4.
l With T,,, greater than or equal to 500*F.
l l
CE-STS 3/4 4-23
o 4
REACTOR COOLANT SYSTEM 3/4.4.9 SPECIFIC ACTIVITY
~
LIMITING CONDITION FOR OPERATION l
1 3.4.9 The specific activity of the primary coolant shall be limited to:
. a.
Less than or equal to 1.0 microcurie / gram DOSE EQUIVALENT I-131, and b.
Less than or equal to 100/I microcuries/ gram.
APPLICABILITY:
MODES 1, 2, 3, 4 and 5.
ACTION:
MODES 1, 2 and 3*.
a.
With the specific activity of the primary coolant greater than
'3 1.0 microcurie / gram DOSE EQUIVALENT I-131 for more than 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> during one continuous time interval or exceeding the limit line shown on Figure 3.4-1, be in at least HOT STANDBY with 7"V9 less than 500'F within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
b.
With the specific activity of the primary coolant greater than 100/l microcuries/ gram be in at least HOT STANDBY with T"'8 less than 500*F within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
MODES 1, 2, 3, 4 and 5:
a.
With the specific activity of the primary coolant greater than 1.0 i
microcurie / gram DOSE EQUIVALENT I-131 or greater than 100/l microcuries/
gram perform the sampling and analysis requirements of item 4 a of Table 4.4-4 until the specific activity of the primary coolant is restored to within its limits.
SURVEILLANCE REQUIREMENTS i
4.4.9 The specific activity of the primary coolant shall be determined to be within the limits by performance of the sampling and analysis program of Table 4.4-4.
"With T,yg greater than or equal to 500*F.
B&W-STS 3/* -25
._._.m m__
O i
^
REACTOR COO _LANT SYSTEM 3/4.4.5 SPECIFIC ACTIVITY LIMITING CONDITION FOR OPERATION 3.4.5 The specific activity of the primary coolant shall be limited to:
Less than or equal to 0.2 sierocuries per gram DOSE EQUIVALENT I-131, and a.
'b.
Less than or equal to 100/l sierocuries per gram.
APPLICABILITY:
OPERATIONAL CONDITIONS 1, 2, 3 and 4.
ACTION:
In OPERATIONAL CONDITIONS 1, 2 or 3 with the specific activity of a.
the primary coolant; 1.
Greater than 0.2 microcuries per gram DOSE EQUIVALENT I-131 but less than or equal to 4.0 microcuries per gram DOSE EQUIVALENT J
I-131 for m'sre than 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> during one continuous time interval or greater than 4.0 microcuries per gram DOSE EQUIVALENT I-131, be in at least HOT SHUTDOWN with the main steam line isolation valves closed within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
l 1
2.
Greater than 100/l microcuries per gram, be in at least HOT SHUT-l l
DOWN with the sain steamline isolation valves closed within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
b.
In OPERATIONAL CONDITIONS 1, 2, 3 or 4, with the specific activity of the primary coolant greater than 0.2 microcuries per gram DOSE EQUIVA-LENT I-131 or greater than 100/l microcuries per gram, perfom the sam-pling and analysis requirene m of Item da of Table 4.4.5-1 until the specific activity of the primary coolant is restored to within its limit.
In OPERATIONAL CONDITION 1 or 2, with:
c.
1 1.
THERMAL POWER changed by more than 15% of RATED THERMAL P0MER 1
in one hour", or 2.
The off-gas level, at the SJAE, increased by more than (10,000) microcuries per second in one hour during steady state operation at release rates less than (75,000) microcuries per second, or l
3.
The off-gas level, at the SJAE, increased by more than (15)% in cne hour during steady state operation at release rates greater l
than (75,000) microcuries per second, perfom the sampling and analysis requirements of Item 4b of Table 4.4.5-1 until the specific activity of the prima'ry coolant-is restored to within its limit.
L t
SURVEILLANCE REQUIREMENTS The specific activity of the reactor coolant shall be demonstrated to 4.4.5 be within the limits by performance of the sampling and analysis program of Table 4.4.5-1.
"Not applicable during the startup test program.
3/4 4-15 GE-STS (BWR/4)
REACTOR C00LANT SYSTEM
.)
3 /4. 4. 5 SPECIFIC ACTIVITY LIMITING CONDITION FOR OPERATION 3.4.5 The specific activity of the primary c'oolant shall be limited to:
Less than or equal to 0.2 sierocuries per gram DOSE EQUIVALENT I-131, a.
and b.
Less than or equal to 100/E microcuries per gram.
APPLICABILITY:
OPERATIONAL CONDITIONS 1, 2, 3 and 4.
ACTION:
In OPERATIONAL CONDITIONS 1, 2 or 3 with the specific activity of a.
the primary coolant; 1.
Greater than 0.2 microcuries per gram DOSE EQUIVALENT I-13 but less than or equal to 4.0 microcuries per gram DOSE EQUIVALENT I-131 for more than 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> during one continuous time interval or greater than 4.0 microcuries per gram DOSE EQUIVALENT I-131, be in at least HOT SHUTDOWN with the main steam line isolation valves closed within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
2.
Greater than 100/I microcuries per gram, be in at least NOT SHUT-DOWN with the main steamline isolation valves closed within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
b.
In OPERATIONAL CONDITIONS 1, 2, 3 or 4, with the specific activity of the primary coolant greater than 0.2 microcuries per gram DOSE EQUIVA-LENT I-131 or greater than 100/I microcuries per gram, perform the sam-pling and analysis requirements of Item 4a of Table 4.4.5-1 until the specific activity of the primary coolant is restored to within its limit.
c.
In OPERATIONAL CONDITION 1 or 2, with:
1.
THERMAL POWER changed by more tha-15% of RATED THERMAL POWER in one hour *, or 2.
The off-gas level, at the SJAE, increased by more than (10,000) nicrocuries per second in one hour during steady state operation at release rates less than (75,000) nicrocuries per second, or 3.
The off-gas level, at the SJAE, increased by more than (15)% in one hour during steady state operation at release rates greater than (75,000) nicrocuries per second, perform the sampling and analysis requirements of Item 4b of Table 4.4.5-1 until the specific activity of the prims'ry coolant is restored to within its limit.
(
SURVEILLANCE REQUIREMENTS 4.4.5 The specific activity of the reactor coolant shall be demonstrated to be within the limits by performance of the sampling and analysis program of Table 4.4.5-1.
"Not applicable during the sta/'...
.u;t --.- :-
G9-STS (BWR/5) 3/4 4-1'
J REACTOR COOLANT SYSTEM 3/4.4.5 SPECIFIC ACTIVITY i
LIMITING CONDITION FOR OPERATION 1
3.4.5 The specific activity of the primary coolant shall be limited to:
a.
Less than or equal to 0.2 microcuries per gram DOSE EQUIVALENT I-131, and j
b.
Less than or equal to 100/l microcuries per gram.
APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, 3 and 4.
]
ACTION:
I a.
In OPERATIONAL CONDITIONS 1, 2 or 3 with the specific activity of the primary coolant; 1.
Greater than 0.2 microcuries per gram DOSE EQUIVALENT I-131 but less than or equai to 4.0.microcuries per gram DOSE EQUIVALENT I-131 for more than 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> during one continuous time interval or greater than 4.0 microcuries per gram DOSE EQUIVALENT I-131, be in at least HOT SHUTDOWN with the main steam line isolation valves closed within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
i 2.
Greater than 100/l microcuries per gram, be in at least HOT SHUT-DOWN with the main steamline isolation valves closed within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
b.
In OPERATIONAL CONDITIONS 1, 2, 3 or 4, with the specific activity of the primary coolant greater than 0.2 microcuries per gram DOSE EQUIVALENT I-131 or greater than 100/E microcuries per gram, perform the sampling and analysis requirements of Item 4a of Table 4.4.5-1 until the specific activity of the primary coolant is restored to within its limit.
a c.
In OPERATIONAL CONDITION 1 or 2, with:
1.
THERMAL POWER changed by more than 15% of RATED THERMAL POWER in one hour", or 2.
The off gas level, at the SJAE, increased by more than (10,000)
I microcuries per second in one hour during steacty state operation l
at release rates less than (75,000) microcuries per second, or j
3.
The off gas level, at the SJAE, increased by more than (15)% in one hour during steady state operation at release rates greater 3
i than (75,000) microcuries per second, perform the sampling and analysis requirements of Item 4b of i
Table 4.4.5-1 until the specific activity of the primary coolant-is restored to within its limit.
j SURVEILLANCE REQUIREMENTS 4.4.5 The specific activity of the reactor coolant shall be demonstrated to be within the limits by performance of the sampling and analysis program of Table 4.4.5-1.
"Not apphcable during the startup test program.
GE-ST5 (BWR/6) 3/4 4-14 i
j ADMINISTRATIVE CONTROLS t.
l ANNUAL REPORTS (Add the following to this section)
The results of specific activi[y analysis ih which the primary coolant exceeded the limits of Specification 3.4.8 (W and CE plants).
3.4.9 (B&W plants) or 3.4.5 (GE plants). The following infomation shall be included: (1) Reactor power history starting 48 to the first sample in which the limit was exceeded; (2). hours prior Results of the last isotopic analy::is for radiofodine perfomed prior to ex-ceeding the limit, results of analysis while limit was exceeded and results of one analysis after the radiciodine activity was reduced to less than limit. Each result should include date and time of sampling and the.radiofodine concentrations; (3) Clean-up system flow history starting 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> prior to the first sample in which' the limit was exceeded; (4) Graph of the I-131~ concentration and one other radiciodine isotope concentration in microcuries per gram as a function of time for the duration of the specific activity above the steady-state level; and (5) The time duration when the specific activity of the primary coolant exceeded the radiciodine limit.
STS-ALL PLANTS
' LIST OF RECENTLY ISSUED GENERIC LETTERS
'a GENERIC LETTER NO.
SUBJECT-DATE 85-03 Clarification of Equivalent' control Capacity 1/28/85 For Standby Liquid Control Systems 85-04 Operator Licensing Examinations 1/29/85 85-05 Inadvertent Boron Dilution Events 1/31/85 85-06 Quality Assurance Guidance for ATWS Equipment that is not Safety-Related 4/16/85 85-07 Implementation of Integrated Schedules.
5/02/85 for Plant Modifications l
l 85-08 10 CFR 20.408 -Termination Reports - Format 5/23/85 85-09 Technical Specifications for Generic Letter 83-28, Item 4'.3 5/23/85 85-10 Technical Specifications.for Generic Letter 83-28, Items 4.3 and 4.4 5/23/85 85-11 Completion of Phase II of " Control of 6/28/85 Heavy Loads at Nuclear Power Plants" NUREG-0612 85-12 Implementation of TMI Action Item II.K.3.5,
" Automatic Trip of Reactor Coolant Pumps" 6/28/85
]
85-13 Transmittal of NUREG-1154 Regarding the Davis Besse Loss of Main and Auxiliary 1
Feedwater Event 8/5/85 84-14 Commercial Storage at Power Reactor Sites of Low Level Radioactive Waste not Generated by the Utility 8/1/85 85-15 Information Relating to the Deadlines for Compliance with 10 CFR 50.49, " Environmental Qualification of Electric Equipment Important l
to Safety for Nuclear Power Plants 8/6/85 85-16 High Boron Concentrations 8/23/85 85-17 Availability of Supplements 2 & 3 to i
NUREG-0933, "A Prioritization of Generic Safety Issues 8/J3/85 85-18 Operator Licensing Examinations 9/27/85 85-19 Reporting Requirements on Primary Coolant 9/27/85 Iodine Spikes
___.-_.m_m-_-
1
'i ATTACHf!ENT 4 j
i PHILADELPHIA ELECTRIC. COMPANY-2301 MARKET STREET P.O. BOX 8699 1
PHILADELPHIA A. PA.19101 g
EowaRO e. 9 AuER, JR.
RECflVfD la sie4s.4ooo ggg EUGEME J. DN AOLEY
.....e.v......
6s.u...k
%fR DON ALD DLA8sM EN j
RUDOLPH A. CH0LLEMS
^# **D=4WN eg l
E. C. au R M M A LL T. H. M AMER CORRELL PAUL AMER 9 ACM August 19, 1986 i
4 EDW A RO J. CULLEN, JR.
THOM AS H. MILLER JR.
Mr. Harold R. Denton, Director I
Office of Nuclear Reactor Regulation l
l U. S. Nuclear Regulatory Commission Washington, D.C.
20555 Re: Limerick Generating Station, Unit 1 Docket No. 50-352
Dear Mr. Denton:
j l
Transmitted herewith for filing with the Commission are 3 originals and 19 copies of Philadelphia Electric Company's Application for Amendment of Facility Operating License NPF-39. This-Application requests changes to the Limerick Technical Specifications regarding reports of iodine spikes to q
conform to the NRC model Technical Specifications.
j In accordance with Section 170.12 of the Commission's regulations, there is enclosed Philadelphia Electric Company's check number'46028851 in the amount of $150 to cover the filing fee for this Application.
Very truly yours, y n c7
- f'.%
//
k 9-s I,
s J, v
Eugene /J. Bradley
/
i EJB:pkc j
Enclosures cc: See Attached Service List 0137q
I i
cc: Troy B. Conner, Jr., Esq.
Benjamin H. Vogler, Esq.
Mr. Frank' R. Romano Mr. Robert L. Anthony Ms. Maureen Mulligan Charles W. Elliott, Esq.
Barry M. Hartman, Esq.
Mr. Thomas Gerusky Director, Penna. Emergency Management Agency Angus Love, Esq.
David Wersan, Esq.
Robert J. Sugarman, Esq.
Kathryn S. Lewis, Esq.
j Spence W. Perry.Esq.
Jay M. Gutierrez, Esq.
Atomic Safety 6 Licensing Appeal Board i
Atomic Safety 6 Licensing Board Panel Docket 6 Service Section i
E. M. Kelly 1
Timothy R. S. Campbell 1
1 1
)
i I
I 1
i BEFORE'THE UNITED STATES NUCLEAR HEGULATORY COMMISSION In the Matter of Docket 50-352 PHILADELPHIA ELECTRIC COMPANY APPLICATION FOR AMENDMENT OF FACILITY OPERATING LICENSE-NPF-39 1
l l
Edward G.
Bauer, Jr.
Eugene J. Bradley.
2301 Market Street Philadelphia, Pennsylvania 19101 i
l Attorneys for
)
Philadelphia Electric Company
]
________I_.__
o
.q.
u BEFORE THEf UNITED STATES NUCLEAR REGULATORY-COMMISSION l
.{
In the Matter of Docket.No. 50-352 PHILADELPHIA ELECTRIC COMPANY APPLICATION FOR AMENDMENT l
1 OF J
FACILITY OPERATING LICENSE NPF-39 l
1 Philadelphia Electric Company, Licensee under Facility Operating License NPF-39 for Limerick Generating Station Unit 1,
hereby requests that the Technical Specifications contained in i
i Appendix A to the Limerick operating license be amended as i
i indicated by a vertical bar in the margin of the attached pages 3/4 4-15, B 3/4 4-4 and Page 6-16.
Licensee proposes to incorporate the-NRC model Technical Specifications provided with Generic Letter 85-19 regarding the requirements for reporting of iodine spikes.
The Commission l
Staff has provided guidance for the reporting requirements of 1
iodine spikes and has determined that because of the improved i
quality of nuclear fuel during the past decade, normal reactor
, - _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - = _ _ _ _ - _ _ _ _ _ - _ - _ _
1 coolant iodine activit, levels have boon maintained wall below 1
the regulatory limits now in effect.
i i
Safety significance
)
4 In an effort to eliminate unnecessary Technical Specification requirements, the Commission Staff in Generic Letter No. 85-19, determined that.some of the existing requirements regarding iodine may be eliminated.
The Limerick Technical Specifications require plant shutdown if the coolant iodine activity limits are exceeded for 800 hours0.00926 days <br />0.222 hours <br />0.00132 weeks <br />3.044e-4 months <br /> in any 12-month period.
In addition to providing for elimination of the existing requirements with regard to plant shutdown, Generic Letter No.
85-19 also provides guidance for the reporting requirements of coolant iodine activity levels, changing the reporting i
requirements from a short term Special Report or a Licensee Event l
Report, to an item which is to be included in an annual report, pursuant to the proposed administrative control 6.9.1.6 paragraph (d) on page 6-16.
Further, page 3/4 4-16 has been left intentionally blank in order to maintain the existing pagination throughout the Technical Specifications.
Previously, page 3/4 4-16 continued the action items of paragraph 3.4.5, which Licensee proposes to eliminate.
Also, page 6-16a has been added because of the additional paragraph (d) on page 6-16.
In order to incorporate the additional paragraph it is necessary to add an additional page 16a.
The (a) designation is used in order to maintain consistent pagination throughout the Technical Specifications. '
1 The safety? significance regarding elimination of soms of-the requirements for Iodine is minimal'.
The; existing regulations (Title 10 CFR 50.73 'Section (b)(1)(li)) requires the Licensee to report fuel cladding failures.
The fuel' cladding ~ design improvements have minimized fuel cladding failures which in the past have been the primary cause for high Iodine levels'in the reactor coolant.
Further, the specific Iodine activity limitations'for the reactor coolant are not being changed and with the existing surveillance requirements, adequate assurance is provided to preclude approaching the specific Iodine limits.
Excessive specific activity levels in the reactor coolant would be detected by the existing surveillance requirements in sufficient time to allow the operators to take corrective actions to minimize the activity levels and maintain the Iodine levels below the limits.
I l
Significant Hazards Consideration Determination i
i 1
Examples of amendments that are considered likely, and l
also not likely, to involve Significant Hazards Considerations j
l
^
were provided by the Commission (48 FR 14870) and include example (iv) as an example of relief granted upon demonstration of acceptable operation from an operating restriction that was imposed because acceptable operation had not yet been demonstrated.
The proposed changes to pages 3/4 4-15, 3/4 4-16, l
B 3/4 4-4, 6-16, and 6-16a, have been previously justified, and fit into the category of example (iv) as amendments which are'not likely to involve Significant Hazards Considerations.
In order to support the No Significant Hazards Consideration l i
l
'datorminntion, nacassaiy background supporting information is provided below, along with an evaluation of each of the three standards set forth in 10CFR Section 50.92.
l i
A e
(1)
The proposed changes do not involve a significant' increase in the probability or consequences of an I
accident previously evaluated because during the past decade, the normal coolant iodine activity in Boiling I
Water Reactors has been maintained at very low levels s
due to superior fuel and fuel cladding design along with i
1 improved operating practices that decrease the probability of cladding failures.
Cladding failures l
previously caused periodic high iodine levels which are I
1 no longer evident in BWR reactor coolant.
Further, the-existing regulation, 10CFR Section 50.72(b)(1)(11),
requires that the Nuclear Regulatory Commission be immediately notified of fuel cladding failures that exceed expected values or that are caused by unexpected I
factors.
l (2)
The proposed change does not create the possibility of a j
new or different kind of accident from any previously
)
evaluated, because acceptable operation has been demonstrated over the past decade due to improved design and manufacturing of nuclear fuel, demonstrating the capability that reactor coolant iodine levels are being maintained well below the limit; thereby reducing the coolant iodine levels which were common in the period 1965 to 1975 because of fuel cladding failures.
_4_
' ~ ~ ~
Chinging tho'acporting requirements does not createntha possibility for a new type of accident from'any.
i previously evaluated..
1 I
(3)
The proposed change does not involve a significant' reduction in a margin of safety because surveillance requirements provide adequate assurance that excessive specific activity levels in the reactor coolant.will be.
detected in sufficient time to allow the operators to-take corrective-action...The specific; iodine activ'ity limitations of the' reactor. coolant have not been changed, and with good fuel management, as has been demonstrated in the past, combined with existing i
reporting requirements for fuel cladding failures, should preclude ever approaching the specific iodine activity limit.
q i
Environmental Consideration This amendment concerns the reporting requirements of j
l iodine spikes and incorporates the NRC model technical specifications provided with Generic Letter 85-19; Based on the discussions above in regards to improved fuel cladding design and j
operating experience.over the last decade, the proposed changes to the Technical Specifications involves no significant increase in the amount and no significant change in the types of any' i
effluents that may be released offsite and there is no i
i significant increase in individual or cumulative oc'cupational i,
l
radiation exposures.
.ne propossd changes will not result in'ony decrease in the safe operation of the plant.
The Plant Operations Review Committee and the Nuclear Review Board have reviewed these proposed changes to the Technical Specifications and have concluded that they do not
~
involve unreviewed safety questions or involve Significant Hazards Considerations and will not endanger the health and i
safety of the public.
Respectfully Submitted Philadelphia Electric Company.
f
', /c.E E.-jf Vice President l
1 _ _ _ _ _
m-__---
COMMONWEALTH OF PENNSYLVANIA :
as.
COUNTY OF PHILADELPHIA I
-i S. L. Daltroff, being first duly sworn, deposes and says:
i i
That he is Vice President of Philadelphia Electric Company, the Applicant herein; that he has read the foregoing Application for Amendment of Facility Operating License and knows j
the contents thereof and that the statements and matters set forth therein are true and correct to the best of his knowledge, information and belief.
J
,l f l
l
-i ]'t:! C.,
i Subscribed and sworn to A
beforemethis/fday of M
l? Yb d-. _, / d i
Notary Public 1
MELANIE R. CAMPANELLA
)
Notary Public. Philadelphia, Philadelphia CO.
My Commission Espires February 12.1990 I
_ m ___
a.___
a UNITED STATES OF AMERICA-NUCLEAR REGULATORY Col + FISSION Before the Atomic Safety and Licensing Board-In the Matter of Docket No. 50-352 PHILADELPHIA ELECTRIC COMPANY.
i (Limerick Generating Station, Unit No. 1)
CERTIFICATE OF SERVICE I hereby certify that copies of Philadelphia Electric Company's Application for Amendment of Facility Operating License NPF-39 in' the.
above-captioned matter were served on the following by deposit in the United States mail, first-class postage prepaid on this 19th day of August, 1986.
Kathryn S. Lewis, Esquire Atom'ic Safety 6 Licensing Municipal Services-Building Appeal Board Panel 15th 6 JFK Blvd.
U. S. Nuclear Regulatory Commission Philadelphia, PA 19107 Washington, D.C.
20555' Benjamin H. Vogler, Esquire Robert J. Sugarman, Esquire Counsel for NRC Staff Sugarman 6 Hellegers Office of the Executive Legal Director 16th Floor, City Place U. S. Nuclear Regulatory Commission 101 North' Broad Street Washington, D.C.
20555 Philadelphia, PA 19107 Angus R. Love, Esquire Troy B. Conner, Jr., Esquire Montgomery County Legal Aid Conner 6 Wetterhahn, P.C.
107 E. Main Street 1747 Pennsylvania Avenue, NW i
Norristown, PA 19401 Washington, D.C.
20006 i
i
~
r
_________M_____
)
Docket 6 Service Section Timothy R. S. Campbell, Director U. S. Nuclear Regulatory Commission
. Department of Emergency Services J
Washington, D.C.
20555 - (3 copies) 14 East Biddle Street West Chester, PA 19380 Mr. Robert L. Anthony 103 Vernon Lane, Box 186 Director Moylan, PA -19065 Pennsylvania Emergency Management Agency Basement, Transportation 6 Safety Building
' David Wersan, Esquire Harrisburg, PA 17120 Assistant Consumer Advocate Office of Consumer Advocate Jay M. Gutierrez, Esquire 1425 Strawberry Square U. S. Nuclear Regulatory Commission Harrisburg,-PA 17120 Region l' 631 Park Avenue Atomic Safety 6 Licensing Board Panel King of Prussia, PA 19406 U. S. Nuclear Regulatory Commission Washington, D.C.
20555 Ms. Maureen Mulligan.
Limerick Ecology Action Mr. Frank R. Romano P.O. Box 761 61 Forest Avenue 762 Queen Street Ambler, PA 19002 Pottstown, PA 19464-Barry M. Hartman, Esquire Charles W. Elliott, Esquire Office of General Counsel Counsel for Limerick Ecology Action-P.O. Box 11775 325 N.10th Street Harrisburg, Pennsylvania 17108 Easton, PA 18042 Mr. Thomas Gerusky, Director E. M. Kelly Bureau of Radiation Protection Senior Resident Inspector Department of Environmental Resources U. S. Nuclear Regulatory Commission l
Fulton Bank Building, 5th Floor P.O. Box 47 Third 6 Locust Streets Sanatoga, PA 19464 Harrisburg, PA 17120 Spence W. Perry, Esquire General Counsel l
FEMA, Room 840 l
500 C Street, SW Washington, D.C.
20472 q
'n l
?/
- u.
p,}.
.e I
Eugene J. Bradley Attorney for i
Philadelphia Electric Company 1
2301 Market Street l-Philadelphia, PA 19101 i
)
ntnutun uvuuant 3xsrzm t
J 3/4.4.5 SPECIFIC ACTIVITY LIMITING CONDITION FOR OPERATION.
L 3.4.5 The specific activity of'the primary. coolant.shall be limited to:
a.
Less than or equal to 0.2 microcurie per gram DOSE EQUIVALENT I-131, and b.
Less than or equal to 100/f microcuries per gram.
APPLICABILITY:
OPERATIONAL CONDITIONS 1,'2, 3 and 4.
ACTION:
a.
In OPERATIONAL CONDITION 1, 2, or 3 with the~ specific activity of the primary coolant; l
1.
Greater than 0.2 microcuries per gram DOSE EQUIVALENT I-131 but less than or equal to 4.0 microcuries per. gram DOSE EQUIVALENT.I-131 for more than 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> during one continuous !
time interval or greater than 4.0 microcuries per gram DOSE EQUIVALENT I-131, be in at least HOT SHUTDOWN with the main
)
steam line isolation valves closed within 12. hours.
2.
Greater than 100/E microcuries per gram, be in at.least HOT SHUTDOWN with the main steamline isolation valves closed within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
b.
In OPERATIONAL CONDITIONS 1, 2, 3 or 4, with the specific activity of the primary coolant greater than 0.2 microcuries per gram DOSE,
EQUIVALENT I-131 or greater than 100/F microcuries per gram,-
perform the sampling and analysis requirements of Item 4a of Table; 4.4.5-1 until the specific activity of the primary coolant is 1
l restored to within its limit.
c.
In OPERATIONAL CONDITION 1 or 2, with:
1.
THERMAL POWER changed by more than 15% of RATED THERMAL POWER 9 in one hour, or 2.
The'off-gas level, at the SJAE, increased by more than 10,000 microcuries per second in one hour during steady state i
operation at release rates less than 75,000 microcuries per j
l second, or 3.
The off gas level, at the SJAE, increased by more than.15% in-one hour during steady state operation at release rates greater than 75,000 microcuries per second, 1
perform the sampling and analysis requirements of Item 4b of Table; 4.4.5-1 until the specific-activity of the primary coolant is restored to within its limit.
SURVEILLANCE REQUIREMENTS 4.4.5.The specific activity of the-reactor coolant shall be demonstrated to be within the limits by performance of the sampling and' analysis program of Table 4.4.5-1.
LIMERICK - UNIT 1 3/4 4-15
REACTOR COOLANT SYSTEn BASES i
i 3/4.4.5 SPECIPIC ACTIVITY The limitations on the specific activity of the primary coolant ensure that the 2-hour thyroid and whole body' doses resulting from a main steam line failure outside the containment during steady state operation will not exceed small fractions of the dose guidelines of 10 CFR Part 100.
The values for the limits on specific activity represent interim limits based upon a parametric evaluation by the NRC of typical site locations.
These values are conservative in that specific site j
parameters, such as SITE BOUNDARY location and meteorological conditions, were not considered in this evaluation.
The ACTION statement permitting POWER OPERATION to continue for limited time periods with the primary coolant's specific activity greater than 0.2 microcurie per gram DOSE EQUIVALENT I-131, but less than or equal to 4 microcuries per gram DOSE EQUIVALENT I-131, accommodates possible iodine spiking phenomenon which may occur following changes in THERMAL POWER.
Operation with specific activity levels exceeding 0.2 microcurie per gram DOSE EQUIVALENT I-131 but less than or equal to 4 microcurie per gram DOSE EQUIVALENT I-131 must be restricted since these activity levels increase the 2-hour thyroid dose at the SITE BOUNDARY following a postulated steam line rupture.
Closing the main steam line isolation valves prevents the release of activity to the environs should a steam line rupture occur outside containment.
The surveillance requirements provide adequate assurance that excessive specific activity levels in the reactor coolant will be detected in sufficient time to take corrective action.
3/4.4.6 PRESSURE / TEMPERATURE LIMITS All components in the reactor coolant system are designed to withstand the effects of cyclic loads due to system temperature and pressure changes.
These cyclic loads are introduced by normal load transients, reactor trips, and startup and shutdown operations.
The various categories of load cycles used for design purposes are provided in Section 3.9 of the FSAR.
During startup and shutdown, the rates of temperature and pressure changes are limited so that the maximum specified heatup and cooldown rates are consistent with the design.
assumptions and satisfy the stress limits for cyclic operation.
I LIMERICK - UNIT 1 B3/4 4-4
ADMINISTRATIVE CONTRC-ANNUAL' REPORTS (Continued) dosimeter, thermoluminescent' dosimeter.(TLD), or film badge measurements.
Small. exposures totalling less than 20% of the-individual total dose need not be accounted for.
In the aggregate, at least 80% of'the total whole-body dose' received from external sources should be assigned to specific major work functions; b.
Documentation of all challenges to safety / relief valves; and Any other unit unique reports required on an annual basis; c.
and d.
The results of specific activity' analysis.in which the primary coolant exceeded the limits of Specification;3.4.5.
The following information-shall be included.(1) Reactor power history starting 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> prior to the first' sample in which the limit was exceeded; (2) Results of the last-isotopic analysis for'radiciodine' performed prior to exceeding the limit,'results of analysis while limit was exceeded and results of one analysis after the radiciodine activity was reduced to less than limit.- Each result'should include date and time of sampling and the radioiodine concentrations; (3) Cleanup system flow history starting 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> prior to the first sample in which the limit was exceeded: (4) Graph of the I-131 concentration and one other radiciodine isotope concentration.in microcuries per gram as a function of time for the duration of the specific activity above the steady-state level; and (5) The time duration when the specific activity of the primary coolant exceeded the radiciodine limit.
MONTHLY OPERATING REPORTS 6.9.1.6 Routine reports of operating statistics and shutdown experience, including documentation of all challenges to the main-steam system safety / relief valves, shall be submitted on a monthly basis to the Director, Office of Resource Management, i
U.S. Nuclear Regulatory Commission, Washington, D.C.
20555, with a copy to the Regional Administrator of the Regional Office of the NRC no later than the 15th of each month following the calendar-month covered by the report.
ANNUAL RADIOLOGICAL ENVIRONMENTAL OPERATING REPORT
- 6.9.1.7 Routine Annual Radiological Environmental Operating Reports covering the operation of the unit during the previous calendar year shall be submitted prior to May 1 of each year.
The initial report shall be submitted prior to May 1 of the year following initial criticality.
The Annual Radiological Environmental Operating Reports shall include summaries, interpretations, and an analysis of trends of the results of the radiological environmental surveillance activities for the report period, includin preoperational studies,g a comparison (as appropriate),
with operational controls and previous LIMERICK - UNIT 1 6-16
THIS PAGE LEFT INTENTIONALLY BLANK.
~
LIMERICK - UNIT 1 3/4 4-16 Aug 08 1985
___-_~ - __ - - -
ADMINISTRATIVE CONTROLS ANNUAL REPORTS ^(Continued) environmental surveillance reports and an assessment of the observed impacts of the plant operation on the environment.
The reports shall also include the results of land use censuses required by Specification 3.12.2.
The Annual Radiological Environmental Operating Reports shall include the results of all radiological environmental samples and of all environmental radiation measurements taken during the report period pursuant to the locations specified in the tables.and figures in the OFFSITE DOSE CALCULATION MANUAL, as well as summarized and tabulated results of these analyses and measurements in the format of the table in the Radiological Branch Technical Position, Revision 1, November 1979.
In the event that some individual results are not available for inclusion with the report, the report shall be submitted noting and explaining the reasons for the missing results.
The missing data shall be submitted as soon as possible in a supplementary report.
The reports shall also include the following:
a summary description of the radiological environmental monitoring program; at least two legible maps.**
- A single submittal may be made for a multiple unit station.
- One map shall cover stations near the SITE BOUNDARY; a second shall include the more distant stations.
LIMERICK - UNIT 1 6-16a h
8 e
me e
e
AEA_CH*iENT 6 REACTOR COOLANT SYSTEM y4.4.5 SPECIFIC ACTIVITY LIMITING CONDITION FOR OPERATION 3.4.5 The specific activity of the primary coolant shall be limited to:
Less than or equal to 0.2 microcurie per gram DOSE EQUIVALENT I-131, a.
and b.
Less than or equal to 100/E micrecuries per gram.
APPLICABILITY:
OPERATIONAL CONDITIONS 1, 2, 3, and 4.
1 ACTION:
i In OPERATIONAL CONDITION 1, 2, or 3 with the specific activity of I
a.
the primary coolant; I
1.
Greater than 0.2 microcurie per gram DOSE EQUIVALENT I-131 but less than or equal to 4 microcuries per gram, operation may i
continue for up to 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> provided that the cumulative operating time under these circumstances does not exceed 800 hours0.00926 days <br />0.222 hours <br />0.00132 weeks <br />3.044e-4 months <br /> in any consecutive 12-month perf.od.
With the total cumulative operating time at a primary coolant specific activity greater than 0.2 micro-j A
curie per gram DOSE EQUIVALENT I-131 exceeding 500 hours0.00579 days <br />0.139 hours <br />8.267196e-4 weeks <br />1.9025e-4 months <br /> in any
's./
consecutive 6-month period, prepare and submit'a'Special Report to the Commission pursuant to Specification 6.9.2 within 30 days indicating the number of hours of operation above this limit.
The provisions of Specification 3.0.4 are not applicable.
2.
Greater than 0.2 microcurie per. gram DOSE EQUIVALENT I-131 for more than 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> during one continuous time interval or for more than 800 hours0.00926 days <br />0.222 hours <br />0.00132 weeks <br />3.044e-4 months <br /> cumulative operating time in a consecutive i
12-month period, or_ greater than 4 microcuries per gram, be in at least HOT SHUTDOWN with the main steam line isolation valves closed within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
3.
Greater than 100/E microcuries per gram, be in at least HOT SHUTDOWN with the main steamline isolation valves closed within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
b.
In OPERATIONAL CONDITION 1, 2, 3, or 4, with the specific activity 1
of the primary coolar.t greater than 0.2 microcurie per gram DOSE EQUIVALENT I-131 or greater than 100/l microcuries per gram, perform the sampling and analysis requirements of Item 4.a) of Table 4.4.5-1 l
until the specific activity of the primary coolant is restored to l
within its limit.
A Special Report shall be prepared and submitted to the Commission pursuant to Specification 6.9.2 within 30 days.
This report'shall contain the results of the specific activity analyses and the time duration when the specific activity of the coolant exceeded 0.2 microcurie per gram DOSE EQUIVALENT I-131 together with the following additional information.
LIMERICK - UNIT 1 3/4 4-15
.q i
1 REACTOR COOLANT SYSTEM-1 LIMITING CONDITION FOR OPERATION (Continued) g ACTION:
(Continued) 1 c.
In OPERATIONAL CONDITION 1 or 2, with:
1.
THERMAL POWER changed by more than 15% of RATED THERMAL' POWER l
in 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> *, or-f 2.
The off gas level, at the SJAE, increased by more than'10,000 i
microcuries per'second in I hour during steady-state operation.
at release rates.less than 75,000 microcuries per second,-or 3.
The off gas level, at the SJAE, increased by more than 15%'in l
1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> during steady-state. operation at release rates greater than,75,000 microcuries per second, perform the sampling and analysis requirements.of Item 4.b) of Table 4.4.5-1 until the specific activity of the primary coolant I
is restored to within its limit.
Prepare and submit to the I
[
Commission a Special Report pursuant to Specification 6.9.2 at i
least on:c per 92 days containing the results of the specific i
activitn analysis together with the below additional information for e3Ch'oCCurrenCd.
, - s Additjonal Information 1.
Reactor power history t. tarting 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> prior to:
1 a)
The first sample'fn which the limit was exceeded, and/or 9
' I.
b)
The ThS! MAL POWER or off gas level change.
2.
Fuel burnup by core regiom 3.
Clean-up flow history starting A8 hours prior to:
a)
The first sample. in which the limit was exceeded, and/or TheTHERMALPON.Roroffgaslevel; change.
b) c C,
Off gas level starting 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> prior to:
1 a)
The first sample in which the linii$'was exceeded, and/or f
t)
The THERMAL PO4ti or off gas. level charge.
i 1 :
SUR\\E!.Lt.ANCE REQUIREMENTS
- s 4.4.5 The specific activity of the reactor coolant shall be demonstrated to l
be within the limits by performance of the sampling and analysis program of Table 4.4.5-1.
'/
- Not applicable during the startup test ' program.
LIMERICK - UH1T 1 3/4 4-10'u T
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REACTOR COOLANT SYSTEM 3/4.4.6 PRESSURE / TEMPERATURE LIMITS REACTOR COOLANT SYSTEM LIMITING CONDITION FOR OPERATION 3.4.6.1 The reactor coolant system temperature and pressure shall be limited in accordance with the limit lines shown on Figure 3.4.6.1-1 (1) curves A and A' for hydrostatic or leak testing; (2) curves B and B' for heatup by non-nuclear means, cooldown following a nuclear shutdown and low power PHYSICS TESTS; and (3) curves C and C' for operations with a critical core other than low power l
PHYSICS TESTS, with:
i A maximum heatup of 100 F in any 1-hour period, a.
b.
A maximum cooldown of 100 F in any 1-hour period, A maximum temperature change of less than or equal to 20*F in any c.
1-hour period during inservice hydrostatic and leak testing opera-tions above the heatup and cooldown limit curves, and d.
The reactor vessel flange and head flange temperature greater than j
or equal to 80 F when reactor vessel head bolting studs are under j
tension.
l APPLICABILITY:
At all times.
l ACTION:
)
With any of the above limits exceeded, restore the temperature and/or pressure to within the limits within 30 minutes; perform an engineering evaluation to determine the effects of the out-of-limit condition on the structural integrity of the reactor coolant system; determine that the reactor coolant system remains
)
acceptable for continued operations or be in at least HOT SHUTDOWN within 12 i
hours and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
1 SURVEILLANCE REQUIREMENTS 4.4.6.1.1 During system heatup, cooldown and inservice leak and hydrostatic testing operations, the reactor coolant system temperature and pressure shall be determined to be within the above required heatup and cooldown limits and to the right of the limit lines of Figure 3.4.6'.1-1 curves A and A',
B and B',
or C and C' as applicable, at least once per 30 minutes.
LIMERICK - UNIT 1 3/4 4-18 l
REACTOR COOLANT SYSTEM BASES 3/4.4.5 SPECIFIC ACTIVITY The limitations on the specific activity of the primary coolant ensure i
that the 2-hour thyroid and whole body doses resulting from a main steam line failure outside the containment during steady state operation will not exceed small fractions of the dose guidelines of 10 CFR Part 100.
The values for the limits on specific activity represent interim limits based upon a parametric evaluation by the NRC of typical site locations.
These values are conservative in that specific site parameters, such as SITE BOUNDARY location and meteoro-logical conditions, were not considered in this evaluation.
The ACTION statement permitting POWER OPERATION to continue for limited time periods with the primary coolant's specific activity greater than 0.2 microcurie per gram DOSE EQUIVALENT I-131, but less than or equal to 4 micro-curies per gram DOSE EQUIVALENT I-131, accommodates possible iodine spiking phenomenon which may occur following changes in THERMAL POWER.
Operation with specific activity levels exceeding 0.2 microcurie per gram D0!E EQUIVALENT I-131 but less than or equal to 4 microcuries per gram DOSE EQUIVALENT I-131 must be restricted to no more than 800 hours0.00926 days <br />0.222 hours <br />0.00132 weeks <br />3.044e-4 months <br /> per year, approximately 10%
of the unit's yearly operating time, since these activity levels increase the 2-hour thyroid dose at the SITE BOUNDARY by a factor of up' to 20 following a postulated steam line rupture.
The reporting of cumulative operating time i
over 500 hours0.00579 days <br />0.139 hours <br />8.267196e-4 weeks <br />1.9025e-4 months <br /> in any 6-month consecutive period with greater than 0.2 micro ~
curie per gram DOSE EQUIVALENT I-131 will allow sufficient time for Commission evaluation of the circumstances prior to reaching the 800-hour limit.
i Information obtained on iodine spiking will be used to assess the parameters associated with spiking phenomena.
A reduction.in frequency of l
isotopic analysis following power changes may be permissible if justified by the data obtained.
Closing the main steam line isolation valves prevents the release of activity to the environs should a steam line rupture occur outside containment.
The surveillance requirements provide adequate assurance that excessive specific activity levels in the reactor coolant will be detected in sufficient time to take corrective action.
3/4.4.6 PRESSURE / TEMPERATURE LIMITS All components in the reactor coolant system are designed to withstand the effects of cyclic loads due to system temperature and pressure changes.
These cyclic loads are introduced by normal load transients, reactor trips, and startup and shutdown operations.
The various categories of load cycles used for design purposes are provided in Section 3.9 of the FSAR.
During startup and shutdown, the rates of' temperature and pressure changes are limited so that the maximum specified heatup and cooldown rates are consistent with the design assumptions and' satisfy the stress limits for cyclic operation.
.I LIMERICK - UNIT 1 B 3/4 4-4
ATTACHMENT 6 s
~~~
1 3874118750 EP-101,: Rov. 5 Page 1 of 20 I
k MJR:mla 1
PHILADELPHIA ELECTRIC COMPANY (4
tt.-
LIMERICK GENERATING STATION
/4 M [
EMERGENCY PLAN IMPLEMENTING PROCEDURE EP-101 CLASSIFICATION OF EMERGENCIES 1.0 PARTICIPANTS 1.1 Shift Superintendent or designated alternate shall assume the role of Emergency Director and implement this procedure, until relieved.
l 1.2 Plant Manaaer or designated alternate shall relieve the Emergency Director, assume the role of Emergency Director, and continue implementing this procedure,-
if necessary.
2.0 ACTIONS-IMMEDIATE CAUTION:
IMPLEMENTATION OF THIS PROCEDURE DOES NOT CONSTITUTE IMPLEMENTATION OF THE EMERGENCY PLAN CAUTION:
THE JUDGEMENT OF THE EMERGENCY DIRECTOR IS VITAL IN PROPER CONTROL OF AN EMERGENCY AND TAKES PRECEDENCE OVER GUIDANCE IN THIS EMERGENCY PLAN PROCEDURE e
4 St I.
s' ge e
_____.y i
j l
l
-EP-101, Rev. 5~
l Page 2 of 20 j
j MJR:mla
'j 2.1 Emergency Director shall:
2.1.1.1 Select categories related to station events'or i
conditions.
Page Number i
Hazards to Station Operation 6
l Environmental 7
1 Loss of Power 8'
1 Personnel. Injury
'9 i
Fire 10 i
Radioactive Release 11 Evacuation of Control Room 12 ramage of Fuel 13 Instrument Failure 14
)
Scram Failure 15 l
Boundary Degradation /LOCA 16
)
Unusual Shutdown 18 Loss of Hot or. Cold Shutdown Capability 19 Security 20 2.1.2 Beginning at the indicated page in Appendix EP-101, review the Emergency Action Levels-(EALS) for categories selected.
2.1.3 Classify the event based on the selected category and EALS.
j IF EVENT TRIGGER IS KNOWN TO BE SPURIOUS, DO NOT l
CLASSIFY EVENT I.E., FALSE HI READING, FALSE CHLORINE MONITOR READINGS ETC.
2.1.4 If the most severe events or conditions are classified as an Unusual Event, implement EP-102, j
" Unusual Event Response."
]
2.1.5 If the most severe events or conditions are classified as an Alert, implement'EP-103, " Alert 1
l Response."
~l 2.1.6 If the most severe events or conditions are i
classified as a Site Emergency, implement EP-104,
" Site Emergency Response."
]
2.1.7 If the most severe events or conditions are classified as a General Emergency,. implement EP-105,
" General Emergency Response."
1
- _ _ _ _ _ _ _ _ - _ _. _. _ _. _ _ ~
l:
EP-101, Rev. 5.
.Page 3 of 20
'l' MJR:mla 3.0'
' ACTIONS-FOLLOW-UP.
'3.1 If the event-is' classified as an Unusual Event,Ethe-Emergency Director shall have a written summary'sent to the NRC withinLtwenty-four. hours of~closecut in accordance with EP-106,. Written Summary
' Notification.-
l 3.2 If event'isEclassified as Alert, Site Emergency, or General Emergency, the. Emergency Director shall:
H l-3.2.1 Periodically evaluate'the' event classification as listed _on. attached Appendix EP-101.
Based upon-results of corrective action taken to1 recover from the emergency situation,Eescalate er.de-escal~ ate the emergency classification.
(It.is preferable, but not mandatory,-to obtain concurrence fromLthe Site Emergency' Coordinator and Corporate Headquarters prior to classification reduction).
The NRC.and appropriate off-site authorities shall be informed of the decision to move from:ene emergency class to the-next.
As' appropriate, l
agencies or personnel listed in phone lists;of.
l Appendix 1 of EPs 102, 103, 104, and 105 shall be notified within 15-minutes once the emergency' level is declared.
3.2.2 Have a' written summary sent to the NRC within eight hours of closecut'or de-escalation'of the emergency:
classification in accordance with EP-106, Written Summary Notification.
W 3.2.3 When the emergency has been controlled and the power plant and auxiliaries have :been placed in a safe shutdown condition, only then will a decision be-made as to whether a recovery phase is' justified.-
Enter the recovery phase after.the emergency'or; accident situation is considered no longer.in effect, obtain the concurrence of the Site Emergency Coordinator and the Emergency Support. Officer at Corporate Headquarters as required per EP-410, Recovery. Phase Implementation.- The recovery phase l
is a departure from an emergency situation.
The t.
Site Emergency Coordina17r and yourself should I
evaluate plant operating conditions as well as the in-plant and out-of-plant radiological-conditions' when making this decision. _ Notifications to the-various individuals and agencies-that the recovery phase has been implemented is the responsibilityLof theLSite Emergency Coordinator.
l EP-101', Rev. S' Page 4 of 20 l
MJR:mla 4.0 APPENDICES 4.1 EP-101-1 Hazards to Station Operation
)
4.2 EP-101-2 Environmental l
4.3 EP-101-3 Loss of Power 4.4 EP-101-4 Personnel Injury 4.5 EP-101-5 Fire 4.6 EP-101-6 Radioactive Release
)
i 4.7 EP-101-7 Evacuation of Control Room 4.8 EP-101-8 Damage of Fuel
]
1 4.9 EP-101-9 Instrument Failure j
4.10 EP-101-10 Scram Failure 4.11 EP-101-11 Boundary Degradation /LOCA 4.12 EP-101-12 Unusual Shutdown i
l 4.13 EP-101-13 Loss of Hot or Cold Shutdown Capability l
l 4.14 EP-101-14 Security i
5.0 SUPPORTING INFORMATION 5.1 Purpose The purpose of this procedure is to provide guidelines for classifying an event or. condition into one of four emergency classifications as described in the Emergency Plan.
Additionally this procedure details the method to change from one emergency action. level to another and to' enter the recovery phase, if applicable.
1 5.2 Criteria For Use l
5.2.1 This procedure shall be implemented whenever the l
Shift Superintendent becomes aware of conditions l
which meet or exceed the Emergency Action Levels.in EP-101, Classification Tables.
L' l
l l.
.EP-101, RGv.T 5 Page 5 of.20 l.
MJR:mla 5.3 Special Ecuipment i
None 1
5.4 References l
(
5.4.1 Limerick Generating Station Emergency Plan I
- 5. 4. 2 '
NUREG 0654 Criteria for' Preparation.and Evaluation Rev. 1 of Radiological Emergency Response Plans'and Preparedness in-Support' of Nuclear Power Plants 5.4.3 EP-102 Unusual Event Response
-5.4.4 EP-102 Appendix 1 Unusual Event Notification
'd l
Message l
5.4.5 EP-103 Alert Response l
5.4.6 EP-103 Appendix 1 Alert Notification Message 5.4.7 EP-104-Site Emergency Response l
5.4.8 EP-104 Appendix 1 Site Emergency ~ Notification
~
5.4.9 EP-105 General Emergency Response l
5.4.10 EP-105 Appendix 1 General Emergency Notification l
5.4.11 EP-106 Written Summary Notification 5.4.12 EP-410 Recovery Phase Implementation l
1
l EP-101, Rav. 5 Appandix.EP-101-6 1
RADIOACT!VE RELEASE Page 11 of 20 l.
MJR:mla UNUSUAL EVENT ALERT 4
1.. Report indicates liquid
- 1. Radiological effluents release
{
effluent release exceeds-greater than 0.5 mR/hr at site j
technical specification.
boundary as indicated by an un-l 3.11.1.1 or 3.11.1.2.
controllable release for greater than 20 minutes with:
- 2. Report indicates gaseous a) North stack effluent radiation effluent release exceeds monitor exceeds 1.0N2 uCi/cc or
{
technical specification
]
3.11.2.1 or 3.11.2.2 or b) South stack effluent radiation q
3.11.2.3 monitor exceeds 1.2N2 uCi/cc.
l i
1
)
J l
SITE EMERGENCY GENERAL EMERGENCY
)
- 1. Radiological effluent release
- 1. Radiological effluent release greater than 50 mR/hr at site greater than 500 mR/hr at site boundary as indicated by boundary as indicated by-an an uncontrollable release for uncontrollable release for greater than 20 minutes with:
greater than 20 minutes with:
a) North stack effluent radiation a) North stack effluent radiation monitor exceeds 1.0 uCi/cc.
monitor exceeds 10 uCi/cc.
- 2. Projected whole body dose
- 2. Projected whole body dose i
greater than.1 rem or. thyroid greater than 1 Rem or thyroid dose greater than.5 Rem at or dose greater than 5 Rem at or beyond the site boundary over beyond the site boundary over course of the event utili:ing course cf the event utilizing RMMS procedure calculating RMMS procedure calculating offsite doses.
offsite doses.
t I
. _ _ _. _ _ -. _ ~
l EP-101, Rsv. 5
.Appandix EP-101-8
-Page 13 of 20
,l MJR:mla DAMAGE OF TUEL UNUSUAL EVENT ALERT
- 1. Steam Jet Air Ejector Discharge
- 1. Steam Jet Air Ejector Discharge radiation monitor exceeds radiation monitor exceeds 2.1P4 mR/hr.
2.1P5 mR/hr
- 2. Steam Jet Air Ejector Discharge 2.
I-131 dose equivalent in the radiation monitor has an un-reactor coolant exceeds 300 uCi/g expected increase of 4000 mR/hr from sample and main' steam over 30 minutes.
line high-high radiation with resultant scram.
- 3. I-131 dose equivalent in the
- 3. Spent fuel damage resulting in'a reactor coolant exceeds refueling floor area. ventilation 0.2 uCi/g from sample exhaust monitor alarm, analysis.
- 4. Containment Post LOCA Radiation Monitors greater than IP2 R/hr.
l SITE EMERGENCY GENERAL EMERGENCY
- 1. Major damage to spent fuel:
- 1. Containment Post LOCA Radiation a) Observation of major damage to spent fuel l
b) Water loss below fuel level in spent fuel pool.
l
- 2. Containment Post LOCA Radiation Monitors greater than 1P3 R/hr.
l l
l
^?T^cw e T 7 NUREG-0991 Safety Evaluation Report related to the operation of Limerick Generating Station, Units 1 and 2 Docket Nos. 50-352 and 50-353 Philadelphia Electric Company U.S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulation August 1983 posc,
ua i
Y LIBRARY
i I
l 15 ACCIDENT ANALYSIS Two groups of design-basis events are evaluated in this section:
anticipated operational occurrences and accidents.
For the analysis of events in either group to be acceptable, a conservative model of the reactor must be used, and all appropriate systems whose operation (or postulated misoperation) would affect the event must be included.
Anticipated operational occurrences are expected to occur during the life of the power plant and are analyzed to ensure that they will not cause damage to either the fuel or to the reactor coolant pressure boundary and to ensure that the radiological dose is maintained within 10 CFR 20 guidelines.
Design-basis accidents are not expected to occur, but I
are postulated because their consequences would include the potential for the i
release of significant amounts of radioactive material.
They are analyzed to l
determine the extent of fuel damage expected and to ensure that reactor coolant pressure boundary damage, beyond that assumed initially by the design-basis I
accident, will not occur, and that the radiological dose is maintained within 10 CFR 100 guidelines.
The postulated radiological consequences of design-basis accidents are given in Table 15.1; the atmospheric dispersion values (x/Q) used in accident evaluations are in Table 15.2.
Accidents i
The applicant has calculated the offsite doses resulting from various postulated design-basis accidents to demonstrate the efficiency of the engineered safety features.
The design-basis accidents represent the upper limits of a wide spectrum of accidents that are considered credible.
In its review, the NRC staff independently performed radiological consequence calculations for the design basis loss-of-coolant, fuel handling, main steamline break, and control rod drop accidents.
For loss-of-coolant accidents, the acceptance criteria for the emergency core l
cooling system specified in 10 CFR 50.46 are 1
(1) The peak cladding temperature must remain helaw 2200 F.
(2) Maximum cladding oxidation must nowhere exceed 17% of the total cladding thickness before oxidation.
(3) Total hydrogen generation must not exceed 1% of the hypothetical amount that would be generated if all the metal in the cladding cylinders, excluding the cladding surrounding the plenum volume, were to react.
(4) The core must be maintained in a coolable geometry.
(5) Calculated core temperatures after successful initial operation of tne emergency core cooling system shall be maintained acceptably low, and decay heat shall be removed for the extended period of time required by j
the long-lived radioactivity remaining in the core.
Limerick SER 15-1
i i
1 Table 15.1 Radiological consequences of design-basis accidents l
Exclusion area
- Low population zone **
Postulated Accident 2-hour dose (rems) 8-hour dose (rems)***
Thyroid Whole body Thyroid Whole body l
Main steamline failure outside containment with concomitant iodine spike 4.2 1.3
- 1. 0 1.4 with pre-accident iodine spike' 83 1.3 18
- 1. 4 i
Rod drop accident 0.7 0.1
- 1. 0 0.04 i
Fuel-handling accident 1.3 0.5 0.3 0.1 LOCA Containment leakage O to 2 hr 196 4
43
<1 2 to 8 hr 1
<1 1
8 to 24 hr 1
1 24 to 96 hr 1
1 96 to 720 hr 1
1 ECCS leakage 0 to 2 hr 63
<1 99
<1 2 to 720 hr 25
<1 MSIV leakage O to 2 hr 27
<1 6
1 2 to 720 hr 1
2.8 t
Total LOCA doses 286 5
178 8
i
" Exclusion area boundary (EAB) distance = 731 meters
- Low population zone (LPZ) boundary = 2043 meters
Anticipated Operational Occurrences Anticipated operational occurences are those transients resulting from single equipment failures or single operator errors that might be expected to occur during normal or planned modes of plant operation.
The acceptance criteria for these transients are based on GDC 10, 15, and 20.
GDC 10 specifies tnat the reactor core and associated control and instrumentation systems be cesigned Limerick SER 15-2
i 1
Based on the ab e discussion, the NRC staff concludes tha the Limerick design for mit' ating the consequences of this accident i acceptable.
15.5 Incre e in Reactor Coolant System Inventory l
The tran ent that could cause unplanned addition coolant inventory is the l
inadve ent actuation of the high pressure coolan injection system.
The high press e coolant injection system has a small e ect, because its flow is smal com red to the recirculation flow.
The trans ent has little effect on fuel th mal margins and on reactor system pressu
- 5. 6 Decrease in Reactor Coolant Invento Inadvertent Opening of a BWR Relief Va
)
l Inadvertent safety / relief valve ope ng causes a decrease in react coolant inventory and results in a mild d pressurization event that has o y a slight effect on fuel thermal margins.
hanges in surface heat flux a calculated to be negligible indicating an ' significant chdnge in minimu critical power ratio.
Thus, the transient i found to be acceptable.
The fect of inadvert-ent safety / relief valve ope ng on suppression pool temper ure is treated in Section 6.2.
Radiological Cor. sequent s of the Failure of Small Line Carrying Primary Coolant Outside Containment The applicant has bmitted an analysis for an in rument line failure.
The applicant's analy is indicates that instrument l' es that penetrate the primary containment and arry primary coolant are to b provided with 1/4-inch orifices.
l In addition, e instrument lines terminate i the secondary containment and meet the int t of RG 1.11 and GDC 55 and 5 The Limer'.k instrument line design is s' ilar to that of other BWRs recently i
license The radiological consequence of a failure in an instrument line l
are ap opriately limited by the 1/4 ' ch orifices and by limitations placed i
on t primary coolant activities by the BWR Standard Technical Specifications.
Bas d on the above, the NRC staff neludes that the Limerick design is a eptable.
Main Steamline Failure Outside Containment A guillotine break of one of the four main steamlines is postulated to occur outside the primary containment, downstream of the outermost isolation valve, resulting in mass loss from both ends.of the break. The primary Coolant loss is limited by steamline flow limiters to a maximum release rate of 200't of rated steam flow, and by Technical Specifications that limit the maximum closure time of the MSIVs to 5.5 seconds.
Mass loss from the broken steamline terminates when the MSIVs are fully closed.
The applicant has calculated that 100.000 lb of water and steam would be released to the atmosphere before the M51Vs would close following such an accident.
The NRC staff followed the recommendation of SRP 15.6.4 and conservatively assumed that 140,000 lb of coolant were released.
As specified in SRP 15.6.4, doses were calculated for tne Technical Specification limits for long-term (Case I) and short-term (Case !!) operation, Limerick SER 15-13
o i
as given NUREG-0123, " Standard Technical Specifications for General Electric Boiling Water Reactors." The calculated doses are listed in Table 15.1.
The assumptions used in the staff analysis are listed in Table 15.4 The resultant doses for the short-term operational limit are within the guideline values of 10 CFR 100, and for the long-term operational limit are less than a small frac-tion for the guideline values of 10 CFR 100.
These results meet the acceptance I
criteria of SRP 15.6.4.
l To provide suitable limitations on the offsite radiological consequences of this accident, the NRC staff will require that the 8WR Standard Technical Specification on primary coolant activity be incorporated into the Limerick Technical Specifications.
Table 15.4 Assumptions used to evaluate the main steamline break accident outside containment Parameter Value Mass of primary coolant released before MSIV closure, Ib 140,000 Fraction of iodine in the primary coolant that is released, %
100 l
l Fraction of noble gases released, %
100 i
Primary coolant concentration (dose equivalent I-131),
microcuries per gram Technical specification limit, normal long-term operation
- 0. 2 Technical specification limit, normal short-term operation 4.0 Other assumptions As in RG 1.5 Based on the above, the NRC staff concludes that the design of Limerick is acceptable with respect to controlling the release of fission products follow-ing a postulated design-basis steamline break accident.
Loss-of-Coolany Accident The applica o has selected and analyzed a hypothetical design-basi (LOCA) and has shown at the distances to the exc sion area boundary (EAB) nd the low populati zone boundary (LPZ) are suf icient to provide reason le assurance that t radiological consequences o such an accident are wit n the exposure guide nes of 10 CFR 100.11(a)(1) d (2).
The analysis has neleded the fol wing sources and radioactive y transport paths to the omosphere:
) contribution from contai ent leakage (2) contribution from pos LOCA leakage from engineer d safety features outside containment (3) contribution fror main steam isolation valve eakage Limerick SER 15-14 I
i CT
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8 l
UNITED STATES OF. AMERICA ~
l NUCLEAR REGULATORY COMMISSION-Before the Atomic Safety'and Licensing Board In the Matter of
)
Philadelphia Electric-Company
)s Docket'No. 50-352-OLA-
)
(TS Iodine)-
(Limerick Generating Station,-
)
Unit 1)
)
LICENSEE'S'STATEMINT OF-MATERIAL FACTS AS TO WHICH THERE IS NO GENUINE ISSUE TO BE HEARD-1.
The amendment proposes no modification to the-Limerick Generating Station radioactive effluent release i
limits.
2.
The. amendment proposes no modification to the J
Station reporting requirements related to plant radioactive effluents.
3.
High levels of iodine in the reactor coolant-encountered by reactors operating in the early 1970's j
resulted from moisture trapped-inside the fuel
- rod, I
pellet-clad interactions, and crud-induced corrosion..
)
i 4.
Improvements in the design of the nuclear fuel, j
1 improved fuel management practices, and'the replacement of j
i l
the older fuel assemblies gradually eliminated the failed' fuel and the resulting higher levels of iodine in operating reactors.
j 1
_ _ _ __a
.(
.I
^ l, i
5.
Since 'startup,- for 'the first operating. cycle,
.j
-5 Limerick has~ averaged only 8 x 10 microcurie per gram.of iodine in the coolant.
l 6.
The average measured value of iodine in the coolant o
at Limerick is 0.04%
of the threshold. value of. 0.2 l
i microcurie per gram contained in the Technical Specifica-j i
tions.
j 7.
Since startup, the maximum value of iodine measured.
)
~4 in the reactor coolant has been 1.2 x 10 microcurie 'per-gram.
I 8.
.The boiling water reactor 1986 median value for
-3 iodine coolant activity was-1.5 x 10 microcurie per gram.
l 9.
Sampling for iodine coolant activity is conducted-at the Station in accordance with Technical Specification 1
4.4.5.
1
)
l 10.
During operation at Limerick, the frequency 'of iodine sampling is daily.
11.
The Station has established an administrative limit l
of 0.02 microcurie per gram which is 1% of the Technical l
l Specification limit.
l l
12.
If the administrative limit for iodine levels in l
the reactor coolant were exceeded, this information would be discussed at the - daily chemistry meetino held at the Sta-tion, management notified, and available courses of action l
considered.
I 13.
The Director, Nuclear Plant Chemistry,= reviews' j
J reactor coolant iodine monitoring data monthly for trends.-
1
1
\\
!]
_3_
4 14.
The NRC has assigned Resident Inspectors to monitor i
operation of Limerick Unit 1.
I i
15.
Periodic inspection reports by the Resident Inspec-l l
tors and by Regional Specialists which include consideration
(
of reactor chemistry are forwarded to Region I and headquar-l ters and are made public.
I 16.
10 C.F.R.
S50.73 (a) (2) (i) requires that a Licensee 1
I Event Report
("LER")
be filed should the iodine coolant activity exceed 4 microcuries per gram, or 0.2 microcurie per gram for 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.
17.
10 C.F.R.
550.73(b) requires that any LER submitted l
l must include the details surrounding the event, its cause l
l and corrective actions and provide a reference to previous l
1 similar events.
18.
LER's related to Limerick Generating Station are placed in the Public Document Room in Washington, D.C.
and I
the Local Public Document Room in Pottstown, Pennsylvania.
l 19.
10 C.F.R.
550.72 (b) (1) (i) requires a one hour noti-fication of the NRC Operations Center via dedicated tele-phone should the todine coolant activity exceed 4 microcurie per gram or 0.2 microcurie per gram for 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.
20.
The Station Emergency Plan requires the declaration i
l of an Unusual Event if the -level of iodine in the reactor l
l coolant exceeds 0.2 microcurie per gram.
21.
The declaration of an Unusual Event would require State and local officials to be notified within 15 minutes 1
I 1
I
.s
_.4._
and the NRC Operations. Center to be notified immediately-i thereafter.
22.
The amendment request L does not seek to eliminate l
any Licensee Event Reports required by 10.C.F.R. S50.73.
j 23.
The. amendment. report does not request any change to Technical Specifications related to offsite release limits or the requirements for monitoring, sampling,. or.' reporting of radioactive effluents.
24.
Any radiological release above regulatory or Technical Specifications limits would require the implemen--
tation of the Station Emergency Plan.
I 25.
The dose calculations for the design basis accident
]
~
1 which is controlled by the iodine level in the' coolant,.the main steamline break accident, is unaffected by the proposed change to the Technical Specifications.
Respectfully submitted, CONNER & WETTERHAHN P.C.
Qult1~
Mark J. Wetterhahn Counsel for Philadelphia Electric Company November 23, 1987 1
1
DOEKETED U %RC.
UNITED' STATES OF; AMERICA-T1 MN 24 21:12 NUCLEAR REGULATORY COMMISSION' 0FFICE Ci SisElarv Before the Atomic' Safety and Licensing Board 00CKEllNG + SEiWIC(.
BRANCH In the Matter of
)
).
Philadelphia Electric Company
)
Docket No. 50-352-OLA
)
(TS Iodine)
(Limerick Generating Station,
)
Unit 1)
)
1 l
CERTIFICATE OF SERVICE 1
I hereby certify'that copies of " Licensee's Motion for Summary Disposition," " Licensee's Memorandum in Support of
'Its Motion for Summary Disposition," " Licensee's Statement of Material Facts as'to which There is No Genuine Issue to be Heard" and " Affidavit of John Doering and John S. Wiley.
in Support of Licensee's Motion.for Summary. Disposition,"
all dated November. 23, 1987 in the captioned matter have been served upon the following by deposit in the United States mail this 23rd day of November, 1987:
Sheldon J. Wolfe, Benjamin H. Vogler, Esq.
Chairman Robert M. Weisman, Esq.
Atomic Safety and Counsel.for NRC Staff Licensing Board Panel Office of the General U.S. Nuclear Regulatory Counsel Commission U.S. Nuclear Regulatory Washington, D.C.
20555 Commission Washington,~D.C.
20555 Dr. Richard F. Cole Atomic Safety and Docketing and Service Licensing Board Panel Section U.S. Nuclear Regulatory U.S. Nuclear Regulatory Commission Commission-Washington, D.C.
20555 Washington,-D.C.
20555 Dr. George A. Ferguson Mr. Frank R.. Romano Atomic Safety and 61 Forest Avenue Licensing Board. Panel Ambler, Pennsylvania 19002 U.S. Nuclear Regulatory Commission Mr. Robert L. Anthony Washington, D.C.
20555 106 Vernon Lane,-Box 186-Moylan, PA 19065 Y
j maw J.' Wetterhahn
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