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- . p Op u Q f R 1""If hH A49W MEMORANDUM FOR: D. G. Eisenhut, Director, Division of Licensing, h7R FROM: C. E. Norelius, Director, Division of Project and Resident Programs
SUBJECT:
REQUEST FOR ASSISTANCE - CORRECTION /C IFICATION OF A TECHNICAL SPECIFICATION DISCREPANCY Fo b/ p (,, J LaSalle Technical Specification 3.3.7.10 requires that the liquid radwaste effluent line radioactivity monitor be operable at all times. This monitor provides an indication of liquid radwaste effluent activity as well as an alarm and termination of discharge function on high activity.
The surveillance requirements for this effluent discharEe monitor are contained in Technical Specification 4.3.7.10. These requirements consist of a daily channel check, a source check prior to each radioactive release, a quarterly channel functional test including alarm and trip functions, and a refueling interval channel calibration. By definition, these surveillance requirements verify that if and when the radiation detector is exposed to a source of radioactivity it will indicate the strength of that source and provide alarm and valve closure functions as required.
The design of the liquid radwaste effluent monitoring system at LaSalle is such that the radioactivity detector is not immersed directly in the effluent stream. The detector is instead located in a separate sample line. Flow through the detector is provided by a sample pump (see attached figure). Sample line flow is sensed by a flow element and alarmed at both high and low values in the control room.
With this system configuration, it is possible to satisfactorily complete the required Technical Specification surveillance requirements for the effluent monitor with no flow through the monitor. Further, and :nore importantly, it is possible to complete an unmonitored radioactive liquid effluent discharge if sample flow rate is not verified.
As currently written, the only flow instruments in the effluent pathway for which Technical Specification operability requirements exist are effluent flow rate instruments. The operability of the sample flow detector / alarms is not specified nor do Technical Specification surveillance requirements exist.
8707100095 870701 PDR FOIA WILLI AM87-121 PDR
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D. G. Eisenhut 2- IMR 2 0 E4 It is our view that Technical Specification requirements for sample flow rate and sample - flow alarm functions should be imposed on the licensee.
This_ position is taken notwithstanding existing procedural requirements for pre-discharge effluent sampling and sample flow alarm status checks.
-The significance of an unmonitored discharge forms the basis for the Region III position. Attached to this memorandum is a copy of the
, applicable sections of LaSalle Technical Specifications marked up to identify recommended changes. Proposed changes'to existing Technical Specifications are identified by a bar in the right-hand margin.
It should also be noted that Standard Technical Specifications do not consider the LaSalle design for radioactive liquid effluent monitoring.
-Thus, the possibility exists that this problem is applicable to other facilities.
Your cooperation in resolving this matter expeditiously is appreciated.
Please respond by June 30th.
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C. E. Norelius, Director Division of Project and Resident Programs i
Enclosures- l
- 1. Tech. Spec. Table 3.3.7.10-1
- 2. Drawing cc w/ enclosures:
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- TABLE 3.3.7.10-1 (Continued)
TABLE NOTATION ACTION 100 -
With the number of OPERABLE channels less than required by the Minimum Channels OPERABLE requirement, effluent releases may continue for up to 14 days provided that i prior to initiating a release:
- a. At least two independent samples are' analyzed in accordance with Specification 4.11.1.1.3, and i
- b. At least two technically qualified members of the Facility Staff independently verify the release rate :
calculations and discharge line valving; Otherwise, suspend release of radioactive effluents via this pathway. ,__
ACTION 101 -
With the number of channels OPERABLE .less than required l by the Minimum Channels OPERABLE requirement, effluent
!. releases via-this pathway may continue for up to 30 days '
provided that, at least once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, grab samples are collected and analyzed at a limit of detection of at least 10-7 microcuries/mi or gamma spectrometric analysis.
ACTION 102 -
With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue for up to 30 days provided the flow rate is estimated at least once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> during actual releases. Pump curves for Instru-ment 3a, or for known valve positions for Instrument 3b, may be used to estimate flow.
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y, nsciou sii 789 noosEvstr noAo 5. g 0 & GLEN ELLYN. ILLINOIS 60137 March 30, 1984 MEMORANDUM FOR: F. Pagano, Jr. Chief I Emergency Preparedness Branch DEPER, 0IE q
FROM: C. J. Paperiello, Chief
- Emergency Preparedness and Radiological Safety Branch, Region III
SUBJECT:
INTERPRETATIONOFPARAGRAPH50.72(b)(1)(v) of 10 CFR PART 50 (F03007984) i With the implementation of the imediate notification requirements in 10 CFR 50.72, the regulation was modified to allow the NRC to eliminate the Unusual Event Classification. However, for events of significance currently classifed by most licensees as Unusual Events a new section((50.72(b)(1)) was added to the regulations. Paragraph 50.72(b)(1)(v) of this new section states as follows:
Any event that results in a major loss of emergency assessment capability, ;
offsite response capability, or communications capability (e.g., significant portion of control room indications, Emergency Notification System, or s offsite notification system).
In regards to this section, our office has received several telephone requests from licensees and the enclosed memorandum from the Kewaunee Senior Resident Inspector requesting clarification of this paragraph under the following general headings:
- 1. What does the term " major loss" mean, especially as it related to the offsite notification system?
- 2. Does the loss of response capability apply only to licensee personnel? ;
- 3. What is considered a " major loss" of offsite response capability and !
how does a licensee or inspector determine that such a loss has or had occurred?
- 4. What does the term "significant portion" of control room indications mean?
- 5. At what point does adverse weather (e.g. severe snowfall or ice storm) indicate a major loss of response, capability for licensee personnel?
We feel that headquarters guidance delineating the meaning of these terms is needed, and should probably take the form of an Information Notice due to its generic implications.
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'F. Pagano March 30; 1984
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In the interim, we have told licensees that further guidance is pending; however, we would interpret the answers to the above mentioned questions as follows: ,
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- 1. A " major loss" means either greater than 50% of the parameter indication, sirens, radios, dedicated communication link, .etc.; or the inability to 1 j
notify greater than 50% of the public in the EPZ due to prompt notification system failure or inoperability. l
- 2. The loss of response capability applies only to the licensee, since the Statement of Considerations related to this paragraph begins by stating
" ... covers those events that would impair a licensee's ability to deal with an accident or emergency." ]
- 3. A major loss of offsite response capability would mean the loss of greater than 50% of the ability of an Emergency Response Facility to function, e.g.,
loss of dose assessment capability, communication links, ability to activate -
an ERF, etc.
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- 4. Significant portion of control room indications means the inability to deter-mine any one particular EAL due to loss of instrumentation needed to assess that EAL. l
- 5. For adverse weather conditions such as heavy. snowfall or ice storms to cause a major loss of response capability for licensee personnel the weather con-dition would have to be of sufficient magnitude to prevent activation of an ,
Emergency Response Facility by any means; e.g., all Emergency Directors l listed would be unable to reach the TSC or E0F, etc. j l
If our interpreation is inconsistent with final guidance, we would like an
, early response informing us of the appropriate final guidance which we will then provide to licensees asking for interpretation of this paragraph.
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- p. C." . Pape'rie lo,. Chief Emergency Prenaredness and Radiological Safety Branch
Enclosure:
Memo frm R. Nelson to N. Jackiw, dtd 1/18/84 cc w/ encl.:
H. Crocker, RI J. Stohr, RII R. Bangart, RIV H. Book, RV Resident Inspectors, RIII 1
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January 18, 1984 MEMORANDUM TOT.: 3.N. J ACT.I W, CH3ET, PROJECTS SECTION 2B TROM: R.L. HELSON, SEN3OR RES3 DENT 3NSPECTOR, KEWAUNEE NUCLEAR POWER PLANT
SUBJECT:
10CPR 50.72 IMMEDIATE NOTIT2 CATION REQUIREMENTS 50:72 (b) (1) states, in part, "..P.the licensee shall notify the NRC as soon as practical and in all cases within one hour of the occurrence of-any of the following:*. 50.72(b) (1) (v) states, in part, "Any event that.results in a major loss of off-site. response capability." On page 50-SC-106 of .the Statements of Consideration, under Paragraph 50.72 (b) (1) (v), it is stated, -
in'part, " Notifying the NRC of these events may permit the NRC to take some compensating measure and to more completely a)sess the consequences of such a loss should it occur during an accident or emergency."
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The above raises three questions for which I am requesting clarification.
(1) Does the loss of response capability apply only to licensee personnel?
(1) What is considered 'a major loss? and (3} How does an inspector or the
~1icensee detern.ine that a major loss has or had occurred? It is fairly common knowledge that 6"-12" snowfalls with drif ting conditions, ice sic:-- etc.
J are not. uncommon occurrences north of the "Masen-Dixon Line" and t) a of conditions have and will continue to occasionally reduce responst c ...lity.
It would seem that adverse weather and/or road conditions would be the most common type of occurrence to reduce off-site response capability and would have been a significant part of the basis fer initiation of the rule, and therefore , consideration would 'have been, given to objective enforceability criteria 'i .e. definition of najor loss. ,-
In summary, please provide me with what. constitutes a major loss, if this is given in (X) percent of plant personnel, then provide me with a method to determine what percent can or would have been able to respond under various adverse weat) and road conditions.
l R.L. Nelson 3 Senior Resident Inspector Kewaunee Nuclear Power Plant
'ec: W.P Shafer, RIII
. Norelius, RIII
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File 'Q MEMORANDUM FOR: Charles E. Norelius, Director Division of Project and Resident Programs..
RegionIIIj' FROM: Darrell G. Eisenhut, Director Division of Licensing, NRR
SUBJECT:
TECHNICAL ASSISTANCE TO ESTABLISH A STAFF POSITION ON A TECHNICAL SPECIFICATIONS INTERPRETATION ON PRIMARY CON-TAINMENT ALLOWABLE LEAKAGE RATES (TIA-83-102)
Re: Memorandum from C.E. Norelius to D.G. Eisenhut, Dated October 21, 1983,
Subject:
" Request for Technical Assistance - NRR Position on Technical Specifications Interpretation - Primary Containment Allowable Leakage Rates."
In the referenced memorandum, Region III requested that NRR provide an inter-pretation of the limit on allowable primary containment leakage during plant operation as specified in the Dresden Technical Specifications and the Standard Technical Specifications. The Containment Systems Branch (CSB) has reviewed the matter and provides the following discussion in response to the Region III request.
The request for clarification of the Technical Specifications regarding allow-able primary containment leakage afose from an event that recently occurred at Dresden 2. During power operation, several small holes were discovered in an expansion bellows of a torus-to-drywell vacuum breaker line. The licensee took the appropriate action in shutting down the unit to repair the holes in the primary containment boundary. Subsequent reports indicated that the calculated leakage from the holes, when added to the most recent Type A test measured leakage, resulted in a total leakage of less than La. Since the Technical Specifications for Dresden 2 indicate limits on containment leakage of La and 0.75 La, the question was raised by Region III as to which limit was applicable during normal operation. In an attenpt tn clarify the Dresden 2 specified limits, Region III consulted the Standard Technical Specifications; however, the same question of interpretation arose. j f
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Charles E. Morelius April 4. 1984 i
The following excerpted sections of the Technical Specifications for Dresden Units 2 and 3 define the limiting conditions for operation in relation to ]!
primary containment allowable leakage:
3.7.2.a(3) "The maximum allowable leakage rate at a pressure of Pa, La, is equal to 1.6 percent by weight of the contaiment air per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> at 48 psig."
3.7.2.b "When prinary containment integrity is required, primary j containment leakage rates shall be limited to: 1 (1) An overall integrated leakage rate for Type A tests of:
(a) Lam less than or equal to 75 percent to La.
(2) (a) A combined leakage rate of less than or equal to 60 percent of La for all testable penetrations and isolation valves subject to Type B and C tests except for nain stean isolation valves."
The staff interpretation of these specifications is as follows-Section 3.7.2.a(3) specifies La as the maximum allowable containment integrated leak rate during normal operation. Section3.7.3.b(2) specifies the additional limit of 0.60 La as the allowable leakage for all Type B and C tested components, except for main steam lines, during normal operation. The third limit of 0.75 La, specified in Section 3.7.2.b.(1).(a)., is intended to ensure that a successful Type A test has been conducted prior to the resumption of peer operations.
Therefore, we agree with the licensee's interpretation that La is the limit on overall integrated leak rate during normal operation. How-ever, in reference to the recent event at Dresden 2, it is the staff's.
position that the expansion bellows are a Type B testable penetration, and therefore, an additional LC0 was applicable. That LCO was the limit of 0.60 La as specified in Section 3.7.2.b(2)(a).
The staff also reviewed the applicable portions of the Standard Technical Specifications to determine the limits on primary containment allowable leakage during nomal operation. All of the related provisions of the Standard Technical Specifications define similar limits on containment leakage and the staff interpretation of these limits is essentially the sane as. for Dresden, Units 2 & 3. Section 3.6.1.2 of the Standard Technical Specifications defines La as the limit on overall integrated leakage rate and 0.60 La as the limit on combined leakage rate for all Type B and C testable penetrations and valves. These limits apply during all modes of nornal operation. The quantity La is a plant-specific de-sign basis leakage that is used in plant accident analyses.
The ACTION statement of Section 3.6.1.2 requires that the " measured" overall integrated containment leakage rates be less than or equal to 0.75 La, prior to resuming power operations. The bases indicate that this limit, as for Dresden 2, is intended to require that a successful
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April 4,1984 Charles E. Norelius Type A test is Itconducted, prior to resuming power is our view that the ACTION statement in during plant operation.Section 3.6.1.2 of the Standard SinceTechnical Specif ;
.to apply only to Type A tests and not during normal operation.
this is not clearly stated in the Technical Specifications, Region III We plan to recomend a revision to this thepoint.
ACTION sta 3.6.1.2 of the Standard Technical Specifications , clarifying In summary, we find that the limit for the allowab f '
of the Technical Specifications for Dresden Also, 2 allowable the limit on and in Section 3.6.1.2.a -
the Standard Technical Specifications.
combined leakage for TypeThis B andlimitCis testable penetrations specified in Section and valve La, during all modes of cperation.3.7.2.b.(2)(a) of the These limits on Technical Spe 3.6.1.2.b of the Standard Technical Specifications. As an added contain '
will not exceed the value assumed in the accident analyses.
conservatism, the measured overall integrated le test.
pated degradation due to plant operations subsequent to each leaka This limit is specified in Section 3.7.2.b(1)(a) of t the Standard Technical Specifications.
- A Darrell G. Eisenhut, Director Division of Licensing Office of Nuclear Reactor Regulation cc: P. McKee, IE R. Starostecki, RI J. Olshinski, RII R. Denise, RIV T. Bishop, RV
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