ML20196G092
| ML20196G092 | |
| Person / Time | |
|---|---|
| Issue date: | 03/18/1997 |
| From: | Tripathi R NRC |
| To: | Lainas G, Sheron B NRC |
| Shared Package | |
| ML20196G081 | List: |
| References | |
| NUDOCS 9705130429 | |
| Download: ML20196G092 (8) | |
Text
-
l From:
Raji Tripathi l
To:
WND2.WNP6.GCL. WND2.WNP6.BWS x""'
Date:
3/18/97 10:13am W*'**
Subject:
CRGR ENDORSEMENT T0: Brian Sheron/Gus Lainas, NRR
SUBJECT:
CRGR ENDORSEMENT OF THE GL ON PASS MODIFICATIONS On January 28. 1997, during the 299th meeting, the Committee to Review Generic Requirements (CRGR) reviewed the generic letter titled " Modification of the NRC Staff's Recommendations for the Post-Accident Sampling System." The CRGR made several comments on the technical aspects and the scope of the proposed generic letter.
Specific recommendations were made.
It was also agreed that the Director. NRR. in accordance with.Section IV(b)(x). CRGR Charter.
Revision 6 would forward a letter to the Chairman. CRGR, stating that the public health and safety and the common defense and security would be adequately protected if the proposed PASS modifications were implemented.
Contingent upon the receipt of such a letter, and the staff's satisfactorily incorporating the Committee's comments, the CRGR agreed to endorse the generic letter for issuance for public comment.
Subsequently, an e-mail, dated March 5.1997, from Aashok Thadani. Associate Director, for Technical Review. NRR to CRGR Chairman, affirmed that he represented the office position when transmitting the PASS GL proposing to eliminate certain redundant and marginal requirements.
Thus, satisfying the intent of the above cited CRGR Charter requirement.
The CRGR endorsement is being formally relayed to the staff via this e-mail.
Plese note that the as endorsed PASS genenric letter was included as an attachment to the combined minutes of the CRGR meeting No. 299 and 300, distributed to the members for their approval, by negative consent, on March 14th. The cognizant staff was also provided a copy of the draft minutes.
If you have any questions, please contact me at 415-7584.
cc:CRGR Members Charlie Haughney Jim Shapaker CC:
DFR TWD2.TWP0. JAM 1, WND1.WNP7.CJH. TWD2.TWP8.MRK....
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PDR REVGP NRCCRGR MEETING 299 PDR
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UNITED STATES E
NUCLEAR REGULATORY COMMISSION OFFICE OF NUCLEAR REACTOR REGULATION WASHINGTON, D.C.
20555-0001 February XX, 1997 NRC GENERIC LETTER 97-YY: MODIFICATION OF THE.NRC STAFF'S RECOMMENDATIONS FOR THE POST-ACCIDENT SAMPLING SYSTEM Addressen.
All holders of operating licenses (cxcept those licenses that h=0 been
- ended to a p0;;c cion onl" status) Or construction permit; for nuclear power PeeraM.sgekceptith6seMhWBEdifiedgglil@ijiansnt@l56tf6n16f reagtore Puroose The U.S. Nuclear Regulator {g glic. p p y & KRC stpff Commission-(NRC) is issuing this generic letter to nptify_addresse.es aboutj e s
WOSg ME 13;JgMc gigolUMirfsni5 idifions F.p deff6r th;e post-accideht sampling system (PASS). 'Although^tthbylpay cons.i this generic TbttFr~ forwards a new staff technical position, no specific action or written response is required.
Backaround I
The need for a PASS was one of the findings endorsed by the Commission following the accident at the Three Mile Island (TMI) plant.
The Commission recommended that all licensed plants have the capability of promptly obtaining and analyzin post-accident samples of the reactor coolant and containment atmosphere is thir[i@findjtime57without permitting radiation ex)osure to any individU 1 that exceeds 5 rem fo the whole body or 50 rem to t1e extremities.
Detailed recommendations for the PASS are specified.in Section II.B.3 of NUREG-0737, " Clarification of TMI Action Plan Requirements." The THI-related recommendatic's specified in NUREG-0737 were subsequently incorporated into 10 CFx 50.34(f)(2)(viii).
However, this rule applied only to applications pending at that time (i.e.. Perkins Nuclear Station, Units 1,
- 2. and 3: Allens Creek Nuclear Generating Station. Unit 1: Pebble Springs Nuclear Plant. Units 1 and 2: Black Fox Station. Units 1 and 2: Skagit/Hanford Nuclear Power Project. Units 1 and 2: and Offshore Power Systems).
On March 17, 1982. NRC issued Generic Letter (GL) 82-05 " Post-TMI Requirements," in which NRC requested that licensees establish a firm schedule for implementing post-accident sampling.
On November 1, 1983, NRC issued GL 83-36 and GL 83-37, " Technical Specifications," which provided guidance on how to address post-accident sampling in the Technical Specifications for boiling-
'~
water reactors (BWRs) and pressurized-water reactors (PWRs), respectively.
In GL 83-36 and GL 83-37. NRC indicated that all licensees should establish, implement, and maintain an administrative program that would include training of personnel, procedures for sampling and analyses, and provisions for sampling and analysis equipment. The licensees could elect to reference this
l program in the administrative controls section of the Technical Specifications l
and include its detailed description in the plant operation manuals.
However.
the recommendations described in Section II.B.3 of NUREG-0737 were imposed as requirements for the majority of operating plants through license conditions or by orders. When a licensee is operating a facility with a license-condition or an order requiring a PASS with the features described below, the licensee must file an application to amend the license in order to obtain NRC authorization to operate in accordance with the new staff technical position, 1
as set forth below.
Descriotion cf Circumstances l
The purpose of the PASS is to obtain information that will help operators monitor the reactivity of the core, the flow of coolant inside the reactor vessel, and the conditions existing inside the containment. These conditions include activity levels, concentration of combustible gases. and the presence of substances that could cause corrosion of the reactor components.
This l
information is obtained from analyses performed on samples of the primary coolant, sump water, and the containment atmosphere.
In Section II.B.3 ot NUREG-0737, the staff recommended 11 criteria for performing sampling and
- analyses of the primary coolant, sump water, and containment atmosphere.
t These criteria specify the types of samples to be taken, the sampling times.
and the analyses to be performed.
I As more information became available on the )ost-accident behavior of nuclear i
plants, it was found that some of the data catained by the PASS might not be j
needed, or might be needed only in the later phases of accident mitigation.
The staff recognized that some relaxation in current PASS requirements was, therefore, acceptable.
In 1987, NRC contracted for a reevaluation of the l
l post-accident sampling program with Pacific Northwest Laboratory (PNL).
The results of PNL's study were documented in NUREG/CR-4330. " Review of Light Water Reactor Regulatory Requirements." Several modifications were proposed, i
but none of them resulted in a revision of the oost-accident sampling program recommendations by the NRC staff.
Subsequently, m response to a series of i
questions from the PASS Owners Group and a rcquest from the Combustion i
Engineering (C-E) Owners Group, which was included in its topical report CEN-415 " Modification of Post Accident Sampling System Requirements," Revision 1-l A. the NRC staff provided clarification and specified its position on several PASS recommendations described in NUREG-0737. In a letter to the PASS Owners Group dated May 30, 1990, and in its safety evaluation report on the C-E PASS Owners Group Report, the NRC staff identified several areas in which some changes were acceptable.
The proposed changes included elimination of the need for determining concentration of hydrogen in the containment atmosphere, and elimination of the need for measuring pH and oxygen concentration in the i
i i
l-More recently, the Electric Power Research Institute (EPRI) requested several relaxations in PASS requirements in the context of the staff's review of EPRI's Advanced Light Water Reactor Utility Requirements Documents. The staff informed the Commission of these requests and their bases in SECY-93-087.
In a staff requirements memorandum dated July 21, 1993, the Commission approved.
with some modifications, these requests for relaxation of PASS sampling requirements as applied to advanced light-water reacters (ALWRs) of evolutionary and passive types. The features that determine PASS sampling J
- ~. _
1 requirements for evolutionary and passive ALWRs are very similar to those of currently licensed facilities. Accordingly, the c cdciftlih reb xation:
in the 'chihsssfith PASS requirements for current fEIIitie5*~kessylbd a)propriste if~the bases for allowing the rebxation: hhinpd5 Ere i'dentical to
~
tiose for the ALWRs, which are set forth below. TheapFovbl.ofPASS rebx tion: EhihgEs for operating reactors would not only simplify the o)eration of~the PASS but would also reduce potential radiation exposure of t1e operators without jeopardizing the safety of the plant.
Discussion Addressees may adopt in their Jost-accident sampling programi the fol'cwing rebxation: modifications to tie NUREG-0737 recommendations toncerning PASS that:are described below.' However, these modifications'should be'esaluated for their potential to femase forlan adverse impact on the effectiveness of a, licensee:s emergency plan, Any decrease in the: effectiveness of this plan" requires prior approval by the NRC. in accordance with'10 CFR 50.54(q)/:
1.icensees who have already done~ so or, propose to isplement;these modifications.
under 10 CFR 50,59 are cautioned to engure that_the_ requirements,of,10'CFR "
t 5,0 54(q) has have been met a s
i
-1.
The recommendation for the analysis of hydrogen by PASS in the containment atmos)here may be eliminated. This analysis is not needed if a containment lydrogen monitor, recommended in Item II.F.1 of NUREG.
0737, is installed in the plant.
This instrumentation, which hou'd g reqUffsdit6 be safety grade, will provide adequate capability for ih6nitifridcfthe concentration of hydrogen in the containment atmosphere after an accident.
2.
The recommendation for the analysis of dissolved gases in the reactor coolant in BWRs may be eliminated. There is no need to obtain reactor coolant samples in BWRs, as follows:
If core uncovery is suspected in this type of reactor, the reactor vessel.is depressurized to approximately the pressure within the wetwell and the drywell.
As a j
result, dissolved gases, mainly hydrogen. will be partially released to the containment atmosphere. and measuring the remaining concentrations i
of dissolved gases in the reactor coolant would not provide meaningful information. The amount of dissolved hydrogen in the reactor coolant before depressurization occurs can be determined based upon its concentration in the containment atmosphere.
In the event of an accident in which only a small amount of cladding damage occurs and the reactor vessel is not depressurized, the capability exists to obtain reactor water samples using the process sampling system later in the post-accident phase, when activity levels are reduced.
l 3.
PWR licensees should ensure that the ca ability exists to determine the i
L gross amount of dissolved gasesMaiiiM hidF6gesi in reactor coolant l
samples to satisfy the intent of~It W
.B.3 of~NUREG-0737. The current i
PASS requirements (in Item II.B.3) are such that this determination L
should be completed within 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />. A sampling and analysis to obtain F
this information performed within 24' hours will be of a quality i
comparable to that performed within 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />. This information is usually used later in the )lant recovery phase. Accordingly, completing the :=pling and sn:1ysi: )y 2' hour; into the post accident ph: c will
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BeteFniintT66"may be extended UMto 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> following an accident.
J 4.
TheP655jrecommendationtoanalyzeforchlorideinreactorcoolant samplss'1n PWRs and BWRs may be eliminated.
Obtaining information on chloride concentration in the reactor coolant is required for assessing the need.to take action to minimize the corrosion of reactor components.
Corrosion in this context is a long-term phenomenon and need not be monitored immediately after an accident. These samples may be taken, i
therefore, at a later post-accident phase, when the reactor coolant j
system is at a low pressure and sampling can be accomplished by the process sampling system.
5.
In NUREC 0737 tThe recommended time for measuring activity levels in
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6.
The recommended time for determining the boron concentration in the reactor coolant may be increased from 3 to 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> following an accident if appropriate alternative instrumentation is installed as described below.
A' knowledge of the boron concentration in the reactor coolant is needed to help in evaluating the reactivity of the core and to provide insight for accident mitigation while the accident is still in progress.
However, for plants with neutron flux monitoring instrumentation that complies with the Category I criteria of Regulatory Guide 1.97.
" Instrumentation for Light-Water-Cooled Nuclear Power Plants To Assess Plant and Environs Conditions During and Following an Accident."
information about the reactivity of the core can be obtained without resorting to boron concentration measurements.
For these plants, the determination of boron concentration in the reactor coolant may be completed within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> after an accident.
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Therefore. Oddrc cc: arc advised that they may introduce these PASS modifications into their facilities on a voluntary basis upon determining that the bases for the modifications as discussed above have been satisfied and the requirements of both 10 CFR 50.59 and 10 CFR 50.54(q) have been met.
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Backfit Discussion The actions described in this generic letter are strictly voluntary.
Therefore, the staff has not performed a backfit analysis.
Federal Reaister Notification I
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i Rcy ster (XX 9 xx"xt on YY.YYYY yy. 1995. CO ent v.'cre received fro-Copics of the staff eva'uatier of the 0 cc-^nt have been made avai' 2'e in the NRC Pub'ic Document R00-Bi bs_Nbsp_Tetsdia ft.Etisip_sblic_Tf6mmentisp_ihi5d!
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Paoerwork Reduction Act Statement This generic letter contains information collections that are subject to the Paperwork Reduction Act of 1995 (44 U.S.C. 3501 et seq.). These information collections were approved by the Office of Management and Budget (OMB).
approval number 3150-0011, which expires July 31, 1997.
The public reporting burden for this collection of information is estimated to 1
average 80 hours9.259259e-4 days <br />0.0222 hours <br />1.322751e-4 weeks <br />3.044e-5 months <br /> per response. including the time for reviewing instructions, searching existing data sources. gathering and maintaining the data needed, and completing and reviewing the collection of information.
NRC is seeking public comment on the potential impact of the collection of information contained in the generic letter and on the following issues:
1.
Is the proposed collection of information necessary for the 3 roper performance of the functions of the NRC including whether tie information will have practical utility?
i
-2.
Is the estimate of burden accurate?
e 3.
Is there a way to enhance the quality, utility, and clarity cr' the information to be collected?
4.
How can the burden of the collection of information be minimized.
including the use of automated collection techniques?
l Send coments on any aspect of this collection of information, including I
suggestions for reducing this burden, to the'Information and Records Management Branch. T-6 F33. U.S. Nuclear Regulatory Commission. Washington. DC 20555-0001, and to the Desk Officer. Office of Information and Regulatory l
Affairs. NE0B 10202 (3150-0011). Office of Management and Budget.
Washington, DC 20503.
NRC may not conduct or sponsor, and a person a nuu iequired to respond to. a collection of information unless it displays a currently valid OMB control number.
l This generic letter requires no specific action or written response.
If you have any questions about this matter, please contact the technical contact listed below or the appropriate Office of Nuclear Reactor Regulation (NRR) project manager.
l Thomas T. Martin Director l
Division of Reactor Program Management l
Office of Nuclear Reactor Regulation Technical
Contact:
Krzysztof I. Parczewski, NRR l
(301) 415-2705 l
pms))Internet: kip @nrc. gov Lead Project Manager: Ronald W. Hernan. NRR (301) 415-2010 E[Rin))Internet:rwh@nrc. gov i
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' to the Combined Minutes 'of CRGR-i Meeting No. 299 and 300 Proposed Generic Letter i-Degradation of Control Rod drive Mechanism and Other Vessel Head Penetration" (January 28 and February 4,1997) l Letter from Ralph E. Beedle, Nuclear Energy Institure, to l
Dr. Denwood F. Ross, Acting Chairman,
- CRGR, dated January 23, 1997 l
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