ML20082E466
| ML20082E466 | |
| Person / Time | |
|---|---|
| Site: | Sequoyah  |
| Issue date: | 04/06/1995 |
| From: | TENNESSEE VALLEY AUTHORITY |
| To: | |
| Shared Package | |
| ML20082E452 | List: |
| References | |
| NUDOCS 9504110235 | |
| Download: ML20082E466 (40) | |
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ii ENCLOSURE 1
. PROPOSED TECHNICAL SPECIFICATION CHANGE. 4
^
SEQUOYAH. NUCLEAR PLANT UNITS'l AND 2 DOCKET NOS. 50-327 AND 50-328.
('IVA-SQN-TS-94-15 ) ' ,
LIST OF AFFECTED PAGES P
Unit 1 i
i License Condition.2.0.(23)F Ilail_2 License Condition 2.0.(16)g -t t
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9504110235 950406 7 i
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E o LIC EN.!;'E No. DPR-71' FDR UNIT l (2) At the first outage of sufficient duration but no later than startup following the second refueling outage. TVA shall install the following qualified monitoring instrumentation:
(a) Integrated monitoring assembly which will accomplish particulate, iodine and noble gas monitoring.
. (b) Containment high range radiation monitor.
E c) Containment pressure monitor, d) Containment water level monitor.
L e) Containment hydrogen monitor.
E. Reactor Coolant System Vents (Section 22.3, II.B.1)
At the first outage of sufficient duration but no later than startup following the second refueling outage, TVA shall install reactor coolant system and reactor vessel head highpoint vents that are remotely operable from the control room.
F. Pett Acti&-t 5::: ling (Sectie- 22.3, II.S.3) gg "^ i: :::h::i:M te crerste d----
the 12:::11:a P :: Accid,s
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R39 22:adnuuta 21,1992, 2:==y ; d 10, M u ::5 22,1994. I
-y:: = ch:11 5- ^r---'le r: 12t:: th:2 :::::Q i 11 1,g th: ::::ad : ft: ling cutege.
H. Instruments of Inadequate Core Cooling (Section 22.3, II.F.2)
(1) By January 1,1982, TVA shall install a backup indication for incore thermocouples. This display shall be in the control room and cover the tenperature range of 200 F - 2000 F.
(2) At the first outage of sufficient duration but no'later than startup following the second refueling outage, TVA shall install reactor vessel water level instrumentation which meets NRC requirements.
- 1. Upgrade Emergency Support Facilities (Section 22.3, II. A.1.M (1) At the first outage of sufficient duration but no later than startup following the second refueling outage, TVA shall update the Technical Support Facilities to meet NRC requirements.
Amendment 35 NOTE: Item G was deleted by Amendment 10 (12/31/81) '
LicEN.cE No. .D PR - 79 FM ' UNn" 2.
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- c. Procedures for Verifying Correct Performance of Operating Activ- '
ities (Section 22.2, i.G.6)
Procedures snall be available to verify the adequacy of the operating . activities.
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'd. Control Room Design (Section 22.2, I.D.1)
TVA.shall consider the benefits of . installing data recording and . ,
logging equipment in the control room to correct the deficiencies '
associated with the trending .of important parameters on strip chart recorders used in the control room as part of the Detailed Control Room Design Review. Implementation shall be carried out in accordance with SECY 82-111B.
- e. Training During Low-Power Testing (Section 22.2, I.G.1) -
One experienced operator trained on Unit I low power testing for natural circulation operation shall be assigned to each shift on' '
this facility. This requirement shall remain until TVA submits-a report, and NRC agrees with findings, that an acceptable level of training and experience on Unit 2 has been attained.
- f. Reactor Coolant System Vents (Section 22.2, II.B.1)'
At the first outage of sufficient duration, but no later than startup following first refueling outage, TVA shall install reactor coolant sys_ tem and reactor. vessel head highpoint vents -
that are remotely operable from the control room.
A g Post Accident S=;;1ing (Sectica 22.2, II.O.0)
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c-11m cyr t- r d e r r i'- r a i- their lett::: da:d ::;;c;;1cr 22 :nd 02:=' .
2' , 100 2, Ja . ar, 9 a nd 10, and " ::P 2', I??': .
The cy t r ch-11 5: :pr:51: ;; 12:n th;; ::r::p .;117;in;; R26 a 9e r^ end refucif:;; =::;; .
- h. Hydrogen Control Measures (Section 22.2, II.B.7)
(1) Prior to startup following the first refueling outage, the Commission must confirm that an adequate hydrogen control system for the plant is installed and will perform its intended function in a manner that provides adequate safety margins.
Amendment No. 26 ;
April 24, 1984 l
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% r ENCLOSURE'2 PROPOSED TECHNICAL SPECIFICATION CHANGE SEQUOYAH NUCLEAR' PLANT UNITS 1 AND 2 DOCKET NOS. 50-327 AND'50-328 (TVA-SQN-TS-94-15)
DESCRIPTION AND JUSTIFICATION FOR POSTACCIDENT SAMPLING-(PAS) b B
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W, 4, Description of Change TVA proposes to modify the Sequoyah~ Nuclear Plant (SQN) technical
_ specifications (TSs).to delete License Condition 2.0.(23)F;of Facility-
. License No. DPR-77 for SQN Unit 1 and 2.0.(16)gLof Facility License-No. DPR-79 for SQN Unit-2. 'These license conditions authorized TVA to, operate SQN's installed postaccident sampling-~(PAS) system:as described.
in TVA letters referenced in'the license condition.
Ramann for change TVA-is submitting TS Change 94-15 to update and clarify SQN's PAS commitments.- SQN's PAS conunitments are described in .five TVA letters, which were incorporated into SQN's operating license (OL) in 1984 (refer-to License-Condition 2.0.23[F] for Unit-l'and 2.0.[16]g for Unit 2). ..The TVA. letters describe specific brands' of postaccident sampling. equipment and instrumentation for on-line. analysis. TVA's-proposed change deletes these PAS Program commitments from the OL'and submits for NRC approval, SQN's revised PAS Program that describes new sampling methods. acceptance criteria, and PAS equipment capabilities.- The license conditions-are--
being proposed.for deletion because a separate requirement to control how the PAS is operated is not needed in addition'to the administrative requirements of TS;6.8.4.e and the functional description contained within Final Safety Analysis Report (FSAR) Sections 9.4.10 and 9.5.10.
This change ~would also facilitate future _ changes under the 10 CFR 50.59 process. TVA's proposed change addresses Inspector Follow-up Items (IFIs) 50-327/93-50-04 and 50-328/93-50-04.
Justification for Changg Condition 2.C.(23)F of License No. DPR-77 for Unit 1 and 2.0.(16)g of License No. DPR-79 for Unit 2 authorized TVA to operate SQN's PAS system as described in letters dated November 23'and December 21, 1983, January 9 and 10, 1984, and March 23, 1984.. . These license condition letters describe TVA's conunitments regarding PAS equipment and capabilities. SQN's PAS facilities were installed between 1981 and 1984 and became fully operational in'1984. The facilities were installed to- _
satisfy the requirements of NUREG-0737, Item II.B.3 and Regulatory Guide (RG) 1.97, Revision.2.
TVA proposes to submit for NRC approval a revised PAS Program for SQN..
The. revised program will be contained within SQN's FSAR.and maintained
.under the 10 CFR 50.59 process. The revised program would replace SQN's current-program commitments contained in the license conditions and would establish PAS as an adminiatratively controlled program governed by SQN TS 6.8.4.e and the more detailed FSAR description. SQN TS 6.8.4.'e states the following:
"Ecstanclient Sampling
'A program which will ensure the capability to obtain and analyze reactor coolant, radioactive lodines and particulates in plant gaseous effluents, j l
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, ' and containment' atmosphere samples under accident conditions. The--
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program shall include the'following:-
(1). ; Training of. personnel-
. (ii): Procedures for sampling and analysis (111). - Provision's.for maintenance of sampling and analysis equipment."-
The.above, program requirements,'in conjunction!with.the FSAR functional'
. description,' provide _ adequate programmatic controls.without-the prescriptive controls implied by the license conditions.
w . Enclosure 4 contains a detailed descriptioncof the revised PAS Program.
- In some areas, the revised PAS Program contains relaxations from the-
- criteria contained in NUREG-0737, Item II.B.3 and RG 1.97, Revision 2.
Justification for any relaxations is also provided in Enclosure 4.
Enclosure 5 contains.a copy of the FSAR changes associated with the revised program.
In conclusion, TVA is submitting for NRC approval SQN's revised! PAS
- Program. . The revised PAS Program will be included in SQN's FSAR and will
' be maintained under the 10 CFR 50.59 process. The revised program _
a continuest to satisfy the PAS objectives in NUREG-0737. Item II.B.3 and-RG 1.97, Revision 2. As._a result of the. proposed revision to:SQN's PAS Program, TVA is also submitting a TS change.to delete SQN_'s' PAS license conditions for both units.
Environmental Impact Evaluation The proposed change request does not involve an unreviewed-environmental' ;
question because operation of'SQN Units l'and 2 in'accordance with this' -I change would not 1 1
- 1. Result in a significant increase in any adverse. environmental impact i previously evaluated in the Final Environmental Statement (FES) as. 'l modified by the staff's testimony to the Atomic' Safety'and Licensing ' .)
Board, supplements to the FES, environmental impact appraisals ~,'or.. '
decisions of the Atomic Safety and Licensing Board.
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- 2. ' Result in a significant change in effluents or power levels.
- 3. Result in matters not previously reviewed in.the licensing basis for j SQN that may have a signfficant environmental' impact. J l
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Enclosure'3 iv.
PROPOSED TECHNICAL SPECIFICATION CHANGE SEQUOYAH NUCLEAR PLANT UNITS 1 AND 2 DOCKET NOS. 50-327 AND=50-328-(TVA-SQN-TS-94-15)
DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATION d
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h O: Significant Hazards, Evaluation.
i ~TVA ha's evaluated-the. proposed technical specification (TS) change andL has determined that.it does not represent a significant' hazards l
> ' consideration based on criteria established in 10 CFR 50.92(c).
Operation of Sequoyah Nuclear' Plant (SQN) in'accordance:with the proposed.
amendment will nots 1
- 1. ' Involve a aignificant increase'in the probability:or. consequences of. ,
- an accident previously evaluated.- l 1
The proposed change involves the deletion,of license conditions.that.~ ;
authorized TVA.to. operate SQN's postaccident sampling (PAS) system. ;
-TVA proposed change establishes programmatic.. control of-SQN's; PAS
- system under SQN TS 6.8.4.e and the SQN Final Safety Analysis ,
i Report. Any future changes to SQN's PAS. Program wouldlbe governed by the 10 CFR 50.59 process. -PAS and analysis will continue at SQN through grab sample acquisition and laboratory analysis and will continue to meet the PAS objectives in NUREG-0737, Item II.B.3 and- '
Regulatory Guide 1.97, Revision 2. Accordingly, the proposed change does'not affect the probability or consequences.of an~ accident. ,
- 2. _ Create the possibility of-a new or different kind of accident from- I any previously analyzed. ,
a The proposed change involves improvements in the operational }
reliability of SQN's PAS system by'using more reliable' laboratory 1 analysis methods, reducing sampling personnel radiation dose,'.and e incorporating practical methods'for sample acauisition and analysis.-
Because the proposed change involves license conditions and sampling-methods that are utilized for postaccident recovery, the potential' '
for an unanalyzed accident is not created. Consequently, no new ,
failure modes are introduced. :
- 3. Involve a significant reduction in a margin of safety.
Plant safety margins are established through limiting conditions of operation, limiting safety ~ system settings,'and safety limits. I specified in the TSs.- As a result of.the proposed amendment, there_ :s will be no change to either the physical design of the plant or,to ..,
any of these settings and limits. The proposed changes do not affect the safe operation of SQN. Therefore, there are no changes to any of the margins of safety.
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I t's ENCLOSURE 4.
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,a PROPOSED TECHNICAL SPECIFICATION CHANGE I
SEQUOYAH NUCLEAR PLANT UNITS.1 AND 2 .!
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~ DOCKET NOS. 50-327.AND 50-328- J i
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DESCRIPTION AND JUSTIFICATION FOR' --
SQN'S REVISED POSTACCIDENT SAMPLING-PROGRAM i
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- INTRODUCTION 'i s ,
The SQN; License Conditions 2.C.(23)F'and 2.C!(16)g-(Unit l'and 2 j respectively) contain the: original:TVA commitment'lettersL(refer to i
?~ ' reference' documents "a.through e",of Attachment 4) associated with SQN's _j
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- Postacident Sampling-(PAS) Program.. ,
TVA proposes to: revise SQN'sl PAS Pr,ogram to clarify the sampling and analysis capabilities. .TVA's revised PAS Program'for the Sequoyah :
Nuclear Plant..(SQN)'isisununarized in Attachment 1. For comparison purpose's, s'_ summary of_SQN's current PAS Program is provided in. !
Attachment 2. The proposed changes continue to satisfy the PAS .
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' objectives of NUREG-0737,. Item II.B.3 and Regulatory Guide l(RG) 1.97, Revision 2. The purpose of revising the PAS: Program .is to: i
'(a) Improve' operational reliability of SQN's PAS facilities, 3 (b). Reduce maintenance associated with PAS online-chemistry '
instrumentation, (c) Utilize more reliable-laboratory analysis methods, '!
(d)~ Reduce PAS operator radiation doses, and .j
-(e) ' Incorporate practical methods for meeting the objectives of. q regulatory _ requirements.
Under the revised PAS Program, SQN's PAS facilities will be' dedicated for-grab sample acquisition. The associated online chemistry-instrumentation +
will no longer be maintained and ' utilized. Sample analysis reliability !
will be maintained with the use of laboratory chemistry instrumentation. ,
Eliminating operation of PAS online chemistry instrumentation ^will- ,
contribute to reducing PAS operator radiation doses. Moreover, the ;
online PAS chemistry instrumentation requires extensive maintenance and ,
is' expensive to replace. '
The proposed' changes include relaxation of postaccident sampling / analysis response times,' exemption of'some PAS parameters and-process changes in the PAS Program. The justifications for relaxation and exemption of PAS .i requirements are based on NRC policy issues for advanced'11ght-water 'l reactors described in NRC memorandum SECY-93-087 from James M. Taylor,
. NRC Executive Director for Operations, to NRC Commissioners dated -
April 2, 1993. The following paragraphs provide the PAS criterion from NUREG-0737, Item II.B.3 followed by a description and justification for SQN's revised program.
' CRITERION 1 ,
The licensee shall have the capability to promptly obtain reactor coolant samples and containment atmosphere samples. The combined time allotted i for sampling and analysis should be three hours or less from the time a decision is made to take a sample.
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LA.: DESCRIPTIONz 1
Each SQN reactor unit has a separate PAS facility that is designed to.
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safely obtain, transfer, and dispose of,'as necessary, samples of: !
. reactor < coolant, containment, sump water, and containment atmosphere._ :
.The PAS' facilities are located in the auxiliary building and contain. l
-all the essential equipment to obtain the necessary samples following- -)
.a' loss-of-coolant-accident (LOCA). Each PAS facility.has a reactor ;
coolant sampling system equipped with a: closed: cooling water heat:
_ exchanger to cool the sample as it is' acquired by.the. liquid sample- a panel (LSP).,: Samples are taken from the reactor' coolant hot legs and from:the. containment sump,_when the residual heat removal system 1 ;
(RHRS) is'in the recirculation mode of! operation.- Acquisition of-the.
, containment atmosphere samples is. performed using a. particulate, l iodine, and gas separation system, in conjunction with a containment' '
air sample panel-(CASP). A summary of the sampling capabilities of l each SQN PAS facility is given in the table below: ;
'SQN PAS FACILITY CAPABILITIES SAMPLE DESCRIPTION SAMPLING LOCATION ,
(1) UNDILUTED AND DILUTED REACTOR REACTOR COOLANT SYSTEM (RCS)'
C00IANT HOT LEG LOOPS.'1 OR'3 (2 LOCATIONS)
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'(2) UNDILUTED AND DILUTED STRIPPED RCS. HOT LEG LOOPS 1 OR 3 GAS FROM PRESSURIZED' REACTOR- (2 LOCATIONS) [
COOLANT
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(3) UNDILUTED AND DILUTED CONTAINMENT CONTAINMENT SUMP VIA DISCHARGE j SUMP LIQUID OF RHRS PUMPS, TRAIN A OR B- -l (2. LOCATIONS) U (4) UNDILUTED AND DILUTED CONTAINMENT COMBINED UPPER AND LOWER '
AIMOSPHERE ~ CONTAINMENT COMPARTMENTS-(2 LOCATIONS) .
The LSPs and CASPs have the capability'to purge sample lines before .!
sampling to help assure that representative' samples can be obtained. l These lines can be flushed after the' sampling operations are~ complete to reduce residual radioactivity in the-lines. :i The LSPs and CASPs use shielded carts / casks for collection of reactor 1 coolant and containment. air samples, respectively. Diluted coolant samples are collected by flushing coolant' captured in calibrated i
aliquot or " bite" valves into sample vials using measured volumes of water. The sample vials are then transported in shielded casks to the ,
i laboratory for analysis. . Undiluted and/or diluted gases can be .
stripped from pressurized reactor coolant samples and collected in evacuated serum vials, which are transported to the laboratory for analysis in a shielded container.
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iAliquots of containment-atmosphere lcan be drawn into a' particulate, iodine, and nobleigas sampler.L These sample components can be. <
transported to.the laboratory for. analysis.in a" shielded carrier.
'- Further' dilutions can:be made, if-necessary, in a' shielded fume hood tp 'using gas' syringes..' Containment atmosphere. hydrogen concentrations
-will be.obtained by reading the containment: hydrogen analyzers.
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SQN's revised PAS Program capabilities and sampling / analysis response
. times are specified in Attachment 1. Following a." potential" core-
-damage-accident.. sampling / analysis will be completed for parameters .
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that support accident recovery within the stated response times:after
" stable" core conditions are regained.- A potential-core damage accident is defined using the following criteria for potential!1oss; of fuel clad barrier. ,
(1) Core exit thermocouples indicate greater than or equal to:
700. degrees Fahrenheit (F), or.
(2) Inadequate core cooling (orange on status. tree) or heat sink (red'on status tree), i.e., RHR shutdown cooling not in service,
-or j (3) Reactor vessel level indicates less than 40 percent with no reactor coolant pump running.
Using this methodology, stable core conditions are then' determined-.to: J be regained by the following criteria:'
(1) . Core exit'thermocouples indicate less than 700-degrees F, +
(2) Adequate core cooling and heat sink exists, and (3) No immediate threats'to safety systems required to-maintain core-cooling.
This will ensure that postaccident samples are representative and analytical results are timely and meaningful for supporting accident recovery. ,
B. JUSTIFICATION Relaxation of Samnling/ Analysis Times Justification for relaxation of postaccident sampling / analysis ;
response. times'and exemption of PAS parameters are based on NRC.- :
policy issues for advanced light-water reactors described in NRC j memorandum SECY-93-087 (see reference document [j]'of Attachment 4). :
The postaccident sampling / analysis response times proposed below are. .!
established in ~ relation to when stable core conditions 'are -regained following a potential core damage accident.- This will ensure that ;
postaccident samples will be representative and analytical-results -i will be timely and meaningful for supporting accident recovery.
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A potentia 1' core damage accident is' defined in NUMARC/NESP-007,
' Revision 2, " Methodology'for Development of Emergency Action- ,
' Levels," dated January 1992 (see reference document [k]io f. .
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. Attachment =4). NUMARC/NESP-007, Revision 2 was approved'by NRC in . j cRG 1.101, Revision 3, " Emergency Planning and Preparedness for Nuclear Power Reactors" (see referenceidocument [1].of
,' ' Attachment 4). TVA is' presently implementing'NUMARC/NESP-007
[ emergency action levels into the SQN Radiological Emergency, Preparedness Program. Using this. methodology.. stable core h' - conditions.are then defined such that the potential will not.-
? continue'to exist for loss of fuel clad barrier and propagation.of.
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fuel damage.
- The proposed response times meet the intent.of NUREG-0737, Item II.B.3 and RG 1.97. The postaccident sampling / analyses will be initiated and completed when stable-plant conditions are regained.
t This also eliminates ambiguity involved with making.the decision to.
take a sample postaccident.
RcS and/or Contain=.nt Si==p Boron Sample and Annivze Within Einht.
Hours After Remainine Stable Core Conditions Followine a Potential ;
f Core P-- ee Accident- .I E .The purpose of-the boron analysis is to prevent criticality during a-degraded core accident. After plant shutdown or reactor: trip, neutron flux detectors are available to continuously monitor core reactivity. SQN's neutron' flux instrumentation complies with the Category 1 criteria of Regulatory Guide 1.97, (i.e...has fully qualified, redundant channels that have' capability to monitor power in the range from 10E-8 percent to 200 percent power. TVA considers the information provided by SQN's neutron flux instrumentation to be acceptable'for monitoring core conditions postaccident.
Accordingly, the RCS will be. sampled'and analyzed.for boron content-within eight hours following a potential core damage accident after regaining stable core conditions.. If a LOCA occurs, the containment sump will be sampled via the RHRS.and analyzed for boron content' within eight hours after stable core conditions are regained following a potential core damage accident. This sampling capability.will ensure that shutdown margin is available:in support.
of accident recovery.
RCS and/or ContA inannt Sinnp p Spectriun - Sa= ale and Analyze Within 24 Hours After Regaining Stable Core Conditions Followine a-fatential Core n===re Accident The gamma spectrum analysis-is used to estimate core damage and to-determine general. plant radioactivity. levels. Core exit .
thermocouples can be used to evaluate core condition'by continuously g- monitoring core exit temperatures. If a LOCA occurs, containment' radiation monitors and containment hydrogen analyzers are also available'to assess core damage. The RCS and/or containment sump (via RHRS) will be sampled and analyzed for gamma emitting radionuclide content within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after stable core conditions are regained following a potential core damage accident. This capability will ensure that the extent of core damage and general plant radioactivity levels are evaluated promptly in support of accident recovery..
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-RCSand/orContain==ntSia=n Gross' Activity - n==nle and An=1vme- .!
}Within'24: Hours'After'Reeminine Stable Core Conditions Followine a -l v V. ,-
Potential Core P-- =e' Accident-j
, ;The gross' activity. analysis-is used to estimate core damage'and. ,
" ~ I determine general ~ plant radioactivity: levels.-l Refer to.
-- justification provided-above.for the-RCS.and. containment sumpfganuna ;
M spectrum analysis..
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. .RCS and/or Contal.= i t S =n' Chloride' LN -le Withik 24 Hours and> '
~ Analvme within-96 Hours After-Regain 4ny SF=hle Core Canditions
. Followinn a Potential Core 6 -e Accident '
-4 The. purpose of the chloride' analysis,islto assess:.the coolant.
corrosion potential'and its effect on long-term corrosion oflRCS'andl
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'l containment components during accident recovery. :The sampling and !
& analysis response times are essentially-the same:as NUREG-0737,.
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, l Item II.B.3. 'j RCS Dissolved Hydrogen or Total Gas R==nle'and Analvse Within. :
j G 14 Hours 'Af ter Regainine Stable Core Conditions Followine -a- 4 Potential Core P-- ee' Accident 4
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The' main purpose of'the RC'S dissolved hydrogen analysis is to assess :
whether hydrogen gas concentration in the RCS is large enough.to 3 form gas bubbles during depressurization that could:possibly y interfere with natural recirculation cooling via the steam generators. This analysis is-also performed,to estimate core damage. ~;
(see the justification-above for RCS and containment' sump gamma spectrum response time). -Core exit thermocouples can be used to ,
determine if core temperatures are high'enough for. cladding ;
hydrolysis reactions'to produce'significant quantities of hydrogen a gas. l Reactor vessel' level indicators are available to monitor.
coolant displacement' caused by' hydrogen' gas bubble formation., Gases ,
that accumulate in the reactor vessel head can be vented via the 'i reactor head vent system. ,
1 If a LOCA occurs, SQN's containment hydrogen analyzers can be read to determine the amount of hydrogen gas released from the RCS. ':This ,
information can then be used to estimate. hydrogen gas concentrations in the RCS. . The RCS will be sampled and: analyzed lfor hydrogen
- content within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after stable core conditionsTare regained ::
following a potential core damage accident.. This capability will- '
ensure that hydrogen gas bubble formation is evaluated prior to RCS ~ :,
depressurization in support of accident. recovery.
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l Containment Atmosphere Ga-a Isotopic - S==nle-and Analyze Within'-
24 Hours After Remainina Stable Core Conditions Following a j
' fdtdial Core Damage Accident The purpose of determining the content of gamma emitting radionuclides in the containment atmosphere is to evaluate core conditions, radionuclide source terms in the event of a breach of i
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containment, and general plant contamination levels'. :.High range: r radiation monitors are availablelfor use-in meeting this is, <
requirement. 1The containment:high-level radiation monitor readings; j
.'can be correlated with estimates of core' damage, radion'uclide source' Fy terms in;the containment atmosphere,'and general plant' contamination '
' levels.in containment.' The containmentiatmosphere will be sampled ;
.and analyzed for gamma emitting'radionuclide content within:24 hours- ;
after stable core conditions are regained following aipotential; core r < 1 damage. accident involving a LOCA.. This capability will ensure
' timely verification of the content of gamma emitting radionuclides in the containment atmosphere in support of accident recovery. j 1
Elimination of Hydrogen and Oxvnen Anm1vais on' Containment' Atmosphere Samples-SQN's containment hydrogen analyzers will be.used to determine hydrogen content in tho containment atmosphere. Piese analyzers. ,
continuously monitor _ hydrogen in the containment a aosphere between .
0 concentrations of 0 percent and 10 percent. Containment hydrogen
- recombiners and igniters are,available to ensure that the a containment. hydrogen concentrations do not build up above.the; 4 10 percent upper-range value of,the containment hydrogen analyzers.. A Elimination of the requirement to analyze hydrogen in containment. >
g atmosphere samples has been granted for other operating. plants l (refer to reference documents [j] and [m] of~ Attachment 4).. W
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Frmpling-of the containment atmosphere and analysis for' oxygen. I
. M ent is not required'in NUREG-0737, Item II.B.3.: Oxygen content is included in RG 1.97. The oxygen analysis. requirement is eliminated from the SQN PAS Program. -Containment'. oxygen concentration can be easily estimated by subtracting'the containment' atmosphere hydrogen analyzer reading from.100 percent and multiplying the difference by 20 percent, which is the-typical oxygen concentration in air.
Elindnation of Potential of Hydrogen (oH)' Analyses on RCS and/or Containment Sump Samples Sampling of the RCS and/or Containment Sump and analysis'for pH is not mandatory in NUREG-0737 Item II.B.3,'but it is included in
< RG 1.97. The purpose of the pH analysis is to ensure that alkaline-(basic) conditions. exist in containment'so that the long-term-corrosion of reactor components is minimized and that reevolution of.
volatile radioiodine species.does not occur. However,~this analysis ~
is not needed because of the reliability-of:the.SQN' containment 4 design. . Required quantities of ice containing specified amounts of-sodium tetraborate are used in the ice, condensers to maintain alkaline conditions in the containment following'a.LOCA.
[ NRC approval for elimination'of this requirement has been given to-the Combustion Engineering Owners Group (refer to reference L
El document [m] of Attachment 4).
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- Elimination 'of- Dissolved Oxvnen Angivais on RCS Samnigg '
i~ m Sampling of the RCS'and' analysis:for dissolved oxygen is not +
4 . mandatory in NUREG-0737,l Item II.B.3, but-it is included in '
. RG ; 1. 9 7_. The. purpose.of'the RCS. dissolved oxygen analysis is-to' assess 1ts7 contribution toward long-term corrosion..;However..this-
~
I analysis hasllittle practical value.in support of' accident. .
recovery. . Based on oxygen solubility.in. water at room temperature,-
the dissolved oxygen concentration'in the RCS following an accident' '
",~ would be-expected;to be approximately 8 parts'per million.(ppm).. '
d : Therefore, measured chloride concentrations ,in -the RCS:and/or '
containment sump will be used with this estimated oxygen-concentration 1to assess the overail long-term effect of coolant- -
g- corrosion on'RCS and containment components.
i 7 NRC approval for elimination of this requirement is documented in reference document-[m] of. Attachment-4.
CRITERION 2 The licensee sha11 establish an-onsite radiological-and chemical analysis N capability to provide, within the three-hour timeframe established aboves ;i quantification of the followingt ,
(1) certain radionuclides in the reactor coolant and containment atmosphere that may be. indicators of the degree of core damage- -
(e.g., noble gases, lodines and cesiums, and nonvolatile' isotopes), ,
(2) hydrogen levels in the containment atmosphere, (3) dissolved gases (e.g.. hydrogen), chloride (time allotted for-analysis subject to discussion below), and boron concentration'of -]
liquids, (4) alternately, have inline monitoring capabilities to perform all or part of the above analyses.
A. DESCRIPTION -I j
SQN's revised PAS Program capabilities for sampling and analysis -
response times, ranges, and accuracies are provided in Attachment 1.
.The response time capabilities are established to meet the intent of .;
NUREG-0737, Item II.B.3 (see justifications under criterion 1). The sampling / analysis range capabilities were approved by NRC for SQN's j RG 1.97 program (see reference document'[h] of Attachment 4). The' analysis accuracy capabilities meet regniatory criteria for analytical procedure accuracies presented at the American' Nuclear ;
Society 1983 Meeting by. Frank Witt (see reference. document (1) of. '
Attachment 4). 3 1
(1) The PAS facilities will be used to obtain diluted and undf?uted samples from the RCS, containment sump and containment
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1 Samples will be transported to the onsite laboratory atmosphere. ;
'for. gamma spectroscopy, analysis.c The: laboratory is. equipped with t as a. computerized, gamma spectroscopy system containing multiple 3 '
'high-resolution ganuna detectors. - This. system will- be used ' to '
. quantify noble gas, iodine and' cesium, and nonvolatile radionuclides required in postaccident samples.
-(2) Containment hydrogen analyzers will be used to fulfill the PAS .;
requirement for determining the levels of hydrogen in the: '
containment atmosphere. The. analyzers have a range of 0 percent '
to 10 percent and an accuracy.of fl.5' percent. 0peration of containment hydrogen recombiners and igniters will ensure thatL the containment atmosphere hydrogen concentration does not build up above the 10 percent upper limit of the analyzers.- -
(3) The LSP will be used to obtain diluted and undiluted RCS and containment sump samples. The dissolved hydrogen and boron analyses will be performed in the SQN radiochemical laboratory.
Provisions are in place to perform. chloride analysis at an offsite laboratory.
(4) The SQN PAS Program will no longer utilize online chemistry instrumentation. ,
B. JUSTIFICATION I ReimvatioD of Sampling / Analysis Ranges The current SQN PAS Program sampling / analysis range capabilities are '
listed in Attachment 2. These range capabilities were approved'by NRC in reference document (h) of Attachment 4. SQN's revised PAS Program sampling / analysis range capabilities listed in Attachment 1.have not.-
been changed; however, the PAS containment atmosphere hydrogen and oxygen analyses, RCS dissolved oxygen analysis, and'RCS/ containment sump pH analyses are not included. Justification was previously.
provided under Criterion 1 for elimination for these parameters.
Relaxation of Analysis Accuracies SQN's revised PAS Program analysis accuracles are listed in Attachment 1. The analysis accuracy capabilities meet requirements for analytical procedure accuracies presented in reference document (i) of Attachment 4. Sampling plus analysis' accuracy capabilities are.
- also included. The decreases in accuracies for RCS.and containment sump boron are due to additional uncertainty associated with the sample aliquot or " bite" valves used in obtaining diluted samples.
For comparison, a table of the present and proposed accuracies are provided in Attachment 3. ,
Enlavation of sample /Analva.is Response Times The response time capabilities are established to meet the intent of NUREG-0737, item II.B.3. Refer to previous' justification for sampling / analysis response times under Criterion 1..
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CRLTERION_.3 Reactor coolant and containment atmosphere sampling during postaccident ,
conditions shall not require an isolated auxiliary system (e.g., the letdown system, reactor water cleanup system) to be placed in operation in order tu use the sampling system.
DESCRIPTION AND JUSTIFICAIl0N Sampling of reactor coolant and containment atmosphere during postaccident conditions does not require any isolated auxiliary systems to be placed in service at SQN. Containment isolation valves required for PAS, which are not accessible after an accident meet IEEE Class 1E requirements.
l CRIIERION.4 Pressurized reactor coolant samples are not required if the licensee can quantify the amount of dissolved gases with unpressurized reactor coolant samples. The measurement of either total dissolved gases or H2 gas in reactor coolant samples is considered adequate. Measuring the 02 concentration is recommended, but is not mandatory.
DESCRIPTION AND JUSTIFICATION l 1
The LSP has the capability to obtain pressurized reactor coolant samples.
The normal sampling sequence for dissolved gas analysis is to depressurize the sample, strip the dissolved gases from solution, and collect the I undiluted stripped gases in an evacuated serum vial. The serum vial is then placed in a shielded container and transported to the laboratory for hydrogen analysis via gas chromatography. Analysis of dissolved oxygen in reactor coolant will not be included in the SQN PAS Program. Refer to the previous justification under Criterion 1 for elimination of dissolved ,
oxygen analysis on RCS samples.
CRITERION 5 l The time for a chloride analysis to be performed is dependent upon two factors: (a) if the plant's coolant water is seawater or brackish water, and (b) if there is only a single barrier between primary containment systems and the cooling water. Under both of the above conditions, the licensee shall provide for a chloride analysis within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of the sample being taken. For all other cases, the licensee shall provide for the analysis to be completed within 4 days. The chloride analysis does not have to be done onsite.
1 DESCRIPTION AND JUSTIFICATION l Factors (a) and (b) described in Criterion 5 do not apply to SQN. The reactor coolant and/or containment sump will be sampled within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after stable core conditions are regained following a potential core damage accident. Provisions are in place to perform chloride analysis on
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L( theLaample(s) at-an:offsite: laboratory within 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br /> after stable core
= conditions are regained following a potential core damage accident. g
' Sampling the RCS and/or containment. sump within 24. hours and then-
- analyzing the samples:within 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br /> meets the intent to complete the ,
analysis within 4 days of the time the~ decision is made to take the samples. Refer to the previous justification under Criterion 1 above for relaxing.the chloride sample / analysis: response times.
CRITERION 6-The design basis.for plant equipment for reactor coolant and containment atmosphere sampling and analyses must' assume that it is possible to'obtain and analyze a sample without radiation exposures to any individual-exceeding the criteria'of GDC 19 (Appendix A, 10 CFR Part 50) (i.e., 5. rem whole body, 75 rem extremities). (Note that the design operational review
. criterion was changed from the operational' limits of 10 CFR 20 (NUREG-0578) to the GDC 19 criterion (October 30, 1979, letter from H. R.
Denton to all licensees.)
DESCRIPTION AND JUSTIFICATION The design basis for plant equipment for reactor coolant and containment atmosphere sampling and analysis is consistent with the radiation exposure
~
limits of GDC 19 (Appendix A 10 CFR Part 50). A postaccident design, basis' sampling mission dose was calculated for cach PAS team member utilizing shielding and time-motion analyses when following existing PAS.
procedures. This calculation'showed that the existing PAS Program design-basis sampling mission' doses are within the specified limits.
Elimination of dependence on PAS online chemistry instrumentation will shorten the necessary time for PAS' facility operators to complete their mission. This'should contribute to reducing the postaccident design-basis- ;
sampling mission doses. N CRITERION 7 The analysis of primary coolant samples for boron is required for pressurized water reactors. (Note that Revision 2 of RG 1.97, when issued, will likely.specify the need for primary coolant boron analyses ';
capability at boiling water reactor plants.) l
.j DESCRIPTION AND JUSTIFICATION u
The SQN PAS Program includes the capability for analyzing boron in RCS H and/or containment sump samples. The sample.and analysis response time,.
range, and accuracies are listed in Attachment 1 and discussed in the response to Criterion 2, Item c.
CR1IERION 8 l
If inline monitoring is used for any sampling and analytical capability specified herein, the licensee shall provide back-up sampling through grab samples, and shall demonstrate the capability of analyzing the samples.
Established planning for analysis at offsite facilities is acceptable.
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p Equipment provided for back-up sampling shall be capable 'of providing at-least one: sample per day for seven days following onset of the accident.
and at least one' sample per week until.the accident condition no longer.
. exists. .
.i DESCRIPTION AND JUSTIFICATION The'SQN' PAS Program will no longer utilize online chemistry Instrumentation.. The PAS facility is dedicated to obtaining grab samples, which are' transported to onsite and offsite laboratories for performing the required analyses. Arrangements have been made for use of a
. Department of Transportation shipping cask and a contract.has been established with an offsite laboratory to run required analyses, especially chloride, c
CRITERION 9 The licensee's radiological and chemical sample analysis capability shall include provisions to: l (1) Identify-and quantify the isotopes of the nuclide categories . .;
discussed above to levels corresponding to the source terms given in RCs 1.3 or 1.4 and 1.7.- Where necessary and practicable, the ability to dilute samples to provide capability for measurement and reduction i of personnel exposure should be provided. Sensitivity of.onsite ;
liquid sample analysis capability should be such as to' permit . ,
measurement of nuclide concentration in the range from approximately l 1 uC1/g to 10 Ci/g. j i
(2) Restrict background levels of radiation in the radiological and chemical analysis facility from sources such that the sample analysis ;
will provide results.with an acceptably small error (approximately a l factor of two). This can.be accomplished through the use of sufficient shielding around samples and outside sources,'and by the use of ventilation system design, which will control the presence of airborne radioactivity. 1 DESCRIPTION AND JUSTIFICATION (1) SQN's revised PAS Program includes the sampling and analysis capabilities-to determine gamma emitting radionuclide concentrations ,
in the range of 1 uCi/g to 10 Ci/g.
(2) Gamma isotopic analyses can be performed on postaccident samples within the required accuracy of a factor of two. Additionally, the other required chemical analyses can.be performed on postaccident i samples to meet the required accuracy criteria. These analyses can be performed within the postaccident sampling design-basis mission dose limits. Refer to the previous responses to Criteria 2 and 6 and Attachment 1.
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CRITERION 10i
< Accuracy, range,-and.sensitivityishall:be adequate to provide' pertinent data to the_operador in order--to. describe radiological ~and. chemical' status W '
of the RCSs'.
.- DESCRIPTION AND JUSTIFICATION g
,,s Referfto the previous response to' Criterion 2.
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<< CRITERION 11-
, In the design of the postaccident sampling:and analysis-capability, consideration should be given';to the following items:
(1) Provisions for purging sample lines, for reducing plateout in sample:
lines - for minimizing sample loss or distortion, for ' preventing -
. blockage of sample lines by loose material in the RCS or containment, for appropriate disposal of-the samples, and for flowLrestrictions to limit reactor coolant loss from a rupture of the sample line. . (The :
-postaccident_ reactor coolant and containment atmosphere. samples ~
should be representative of'the reactor coolant in the core area and.
the containment atmosphere following a transient or accident. :The.
sample lines should be as short as possible to minimize the' volume of-fluid to be taken from containment.- The residues of sample-
! collection should.be returned to-containment or'to a closed system.
(2) Theventilationexhaustfromthesamplingstationshould'be[ filtered.
with charcoal absorbers and high-efficiency ' particulate air (HEPA) :
filters.
DESCRIPTION AND JUSTIFICATION (1) .SQN's LSPs and CASPs have the capability to purge sampling lines before. sampling to help assure that representative samples can be obtained. The LSP sampling lines can be flushed with demineralized water after sampling operations are completed to reduce residual radioactivity in the lines'and to reduce particulate and chemical.
species plateout.
To reduce plateout in the containment atmosphere-sampling lines,.the lines are heatLtraced and thermally; insulated. This heat trace-should maintain the sample stream temperature'in route to theLPAS facility, thus reducing iodine plateout and steam condensation, thereby'
. retaining the sample integrity. . The containment atmosphere sampling lines are also flushed with nitrogen to help minimize plateout.
1 A tight system must.be maintained to minimize sample loses'or-distortion. The sampling lines were pressure tested after
. installation. Valves,-(i.e., hand, check, or solenoid) and line welds were~ chosen for their abilities to' minimize fluid leakage.
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p E Following each sample acquisition process, the liquid ~1ines are j backflushed with demineralized yater, and the containment atmosphere
- t. sampling lines are backflushed with nitrogen. These flushing operations clean'out the previous sample and reduce residual q radioactivity in the lines; thereby aid in the prevention of-sample j distortion.
Filters are used to prevent blockage of RCS and containment sump F sampling lines.- Also, pipe scale or crud should be minimized due to ,
the use of stainless steel as the piping and tubing material. With this design configuration, only fine entrained particles should be
- present in the liquid samples.
All samples will be returned back to the disabled unit or to a closed.
system. Each PAS facility liquid' sampling system has in its design a waste collection tank. During tecident conditions, liquid sample waste will be routed from this tank to the disabled reactor unit.
During training exercises, the contents of-the waste tank are routed
- to the radwaste system.
Redundant IEEE Class 1E solenoid operated isolation valves, which trip closed by operator action,'will be used to isolate liquid sample lines and prevent further coolant loss in the event of a sample line rupture.
The postaccident reactor coolant and containment atmosphere samples-should be representative of the reactor coolant in the core area and the containment atmosphere, respectively, following a transient or accident. These. samples are acquired from locations described in the response to Criterion 1.
(2) The PAS facility environmental control system provides ventilation during normal plant operations and training activities. In addition, ventilation and control of airborne radiological contamination is provided during postaccident acquisition and testing of samples.
This is accomplished through pressurization of the areas by the ventilation system, which induces air from areas of lesser to areas 1 of greater contamination potential.
During normal plant operation, ventilation air is supplied to the facility via the auxiliary building general ventilation system and an auxiliary supply fan. Exhaust air is sent directly to the auxiliary building general ventilation system.
During postaccident conditions or sampling operations, the normal ,
supply and exhaust systems are isolated and ventilation air is taken-directly from the outside at a point on the roof of the Unit 1 additional equipment building. Both the Unit 1 and Unit 2 systems share this common intake. A supply fan provides air to the sampling side of the facility in response to a differential pressure controller. Air is drawn by an exhaust fan from both the sample and valve gallery areas through the PAS facility gas fan treatment subsystem, which consists of one HEPA/ charcoal-type air cleanup .
unit. The filtered air is then routed to the exhaust duct downstream of the auxiliary building gas treatment system (ABGTS).
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The PAS facility environmental' control system is not a nuclear' l safety-related-system; however. thel isolation valves.and-duct that--
' interface with the ABGTS and the auxiliary building' secondary
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containment enclosure are designed to Category 1. standards. These valves:are.also backed by'. Class 1E power.. All remaining portions of'
~)~ .the system are designed to Category-1(L). requirements.
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ATTACHMENT 1 ,
SEQUOYAH NUCLEAR PLANT REVISED POSTACCIDENT SAMPLING PROGRAM SAMPLING / ANALYSIS #4ALYSIS SAMPLING / ANALYSIS SAMPLING / ANALYSIS SAMPLE POINT PARAMETER UNITS RANGE
- ACCURACY ACCURACY RESPONSE TIME SAMPLE TYPE RCS AND/OR CONT. BORON PPM 50 to 6000 5% (1000-6000)* 10%'(500-6000)* 8 HOURS 9,# GRAB SAMPLE SUMP ** 50 (50-1000)* 50 (50-500)*
RCS AND/OR CONT. GAMMA uCi/mL ISOTOPIC FACTOR OF TWO FACTOR OF TWO 24 HOURS # GRAB SAMPLE SUMP" SPECTRUM ANALYSIS RCS AND/OR CONT. GROSS uCi/mL 10 to 1E+7 FACTOR OF TWO FACTOR OF TWO 24 HOURS # DETERMINE BY SUMP ** ACTIVITY TOTALING GAMMA ISOTOPIC ACTIVITIES RCS AND/0R CONT. CHLORIDE PPM 0.1 to 20 10% (0.5-20)* 10% (0.5-20)* 24 HOURS (SAMPLING)# PROVISIONS ARE SUMP ** 0.05 (0.1-0.5)* 0.05 (0.1-0.5)* % HOURS (ANALYSIS)# ESTABLISHED FOR OFF SITE ANALYSIS RCS DISSOLVED CC(STP) 10 to 2000 20% (50-2000)* 20% (50-2000)* 24 HOURS # GRAB SAMPLE HYDROGEN OR Kg 5.0 (10-50)* 5 (10-50)*
TOTAL GAS CONTAINMENT GAMMA uCi/CC ISOTOPIC FACTOR OF TWO FACTOR OF TWO 24 HOURS # GRAB SAMPLE ATMOSPHERE SPECTRUM ANALYSIS CONTAINMENT HYDROGEN PERCENT 0 to 10 1.5 1.5 NOT APPLICABLE DETERMINE BY ATMOSPHERE READING CONTAINMENT HYDROGEN ANALYZERS
- THE CONTAINMENT SUMP IS SAMPLED VIA THE RESIDUAL HEAT REMOVAL SYSTEM FOLLOWING A LOSS-OF-COOLANT ACCIDENT.
- THE SAMPLING / ANALYSIS RANGES HAVE BEEN APPROVED AS PART OF THE SEQUOYAH NUCLEAR PLANT REGULATORY GUIDE 1.97 FINALIZED PROGRAM.
ACCURACIES ARE EXPRESSED AS 1 STANDARD DEVIATION (68% CONFIDENCE INTERVAL) UNCERTAINTY ESTIltATES.
@ IF POTENTIAL CORE DAMAGE IS NOT INDICATED FOLLOWING AN ACCIDENT, THE RCS WILL BE SAMPLED AND BORON ANALYZED TO VERIFY SHUTDOWH MARGIN AS SOON AS POSSIBLE IN RESPONSE TO EMERGENCY PROCEDURES.
- FOLLOWING A POTENTIAL CORE DAMAGE ACCIDENT, SAMPLING / ANALYSIS WILL BE COMPLETED WITHIN THE STATED RESPONSE TIMES AFTER REGAINING STABLE CORE CONDITIONS.
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ATTACHMENT.2 ~
.g SEQUOYAH MUCLEAR PLANT CURRENT POSTACCIDENT: SAMPLING ~ PROGRAM ~-
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APPROVED - APPROYED ANALYSIS SAMPLING / ANALYSIS, SAMPLE POINT PARAMETER UNITS RANGE # ACCLRACY RESPONSE TI L SAMLE TYPE RCS AN0/0R CONT. BORON- PPM 50 to 6000 6% (2 SIGMA) 3 HOURS GRAB SAMPLE SUMP **
s RCS AND/0R CONT. -ISOTOPIC uC1/mL ' ISOTOPIC ANALYSIS FACTOR OF TWO 3 HOURS- GRA8 SAMPLE?
SUMP" GAPMA SPECTRUM RCS AND/0R CONT. GROSS uCi/mL 10 to 1E+7 FACTOR OF TWO 3 HOURS. -SUM OF.THE GAfMAS-SUMP" ~ ACTIVITY RCS AND/OR CONT. CHLORIDE PPM 0.1 to 20 20% (1-20) - 24/% HOURS ON-LINE ANALYSIS LE SUMP ** 151 (0.1-1)
RCS AND/OR CONT. pH pH UNITS 1 to 13' i 0.5% 3 HOURS' ON-LINE ANALYSIS SUMP **
RCS AND/OR CONT. DISSOLVED PPM 0.1 to 20 t 10% 3 HOURS ON-LINE ANALYSIS SUMP ** -OXYGEN RCS DISSOLVED CCfSTP) 10 to 2000 15% 3 HOURS ON-LINE' ANALYSIS HYDROGEN OR Kg -
TOTAL GAS.
CONTAINMENT GAPMA vC1/CC ISOTOPIC ANALYSIS FACTOR OF TWO 3 HOURS GRAB SAMPLE-ATMOSPHERE ' SPECTRUM CONTAINMENT HYDROGEN PERCENT 0 to 10 1* 3 HOURS CONTAINMENT HYDROGEN' ATMOSPHERE ANALYZER READING WITH'
.GRA8 SAMPLE CAPABILITY.
CONTAINMENT OXYGEN ' PERCENT -O to 30- NA' 3 HOURS GRABSAMPLEj.
ATMOSPHERE o* THE CONTAINMENT: SUMP IS SAMPLED VIA THE RESIDUAL HEAT REMOVAL SYSTEM FOLLOWING A LOSS-OF-COOLANT ACCIDENT.
O ACCURACY _ WAS PROVIDED FOR CONTAIfetENT HUMt0 GEN ANALYZERS.
o SAMPLING / ANALYSIS CAPA8ILITY IS REQUIRED 3 HOURS AFTER THE DECISION IS MADE TO TAKE SAMPLES, EXCEPT FOR CHLORIDE. CHLORIDE SAMPLING / ANALYSIS IS; REQUIRED 24 HOURS /% HOURS AFTER THE DECISION IS MADE.
O REFER TO RG 1.97_COPMITMENTS IN REFERENCE DOCUMENT'(h) 0F ATTACHMENT 4.
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ATTACHMENT 3 2,
-SEQUOYAH NUCLEAR PLANT POSTACCIDENT SAMPLING PROGRAM ANALYSIS ACCURACY COMPARISON CURRENT REQUIRED PROPOSED 4 PROPOSED TOTAL' ANALYSIS -ANALYSIS ANALYSIS-. SAMPLING / ANALYSIS SAMPLE POINT PARAMETER UNITS ACCURACY 9 ACCURACY # ACCURACY ~ ACCURACY RCS AND/OR CONT. BORON PPM 6% (2 SIGMA) 5% (1000-6C00) 5% (1000-6000)* 31 .10%-(500-6000)*
50 ( 1000)
SUMP ** - 50 (50-1000)* i 50- (50-500)*-
RCS AND/0R CONT. GAMMA uCi/mL FACTOR OF TWO FACTOR OF TWO FACTOR OF TWO FACTOR OF TWO SUMP ** SPECTRUM RCS AND/0R CONT. GROSS uCi/mL FACTOR OF TWO FACTOR OF TWO - FACTOR OF TWO. FACTOR'0F TWO SUMP ** ACTIVITY RCS AND/OR CONT. CHLORIDE PPM i 20% (1-20) 10% (0.5-20) -
10%'(0.5-20)* 't 10% -(0.5-20)*
SUMP ** 15% (0.1-1) i 0.05 ( 0.5) i 0.05 (0.1-0.5)* 't 0.05 (0.1-0.5)*
RCS DISSOLVED CC(STP) 15% 20% (50-2000)- 20% (50-2000)* 20% (50-2000)*
HYDROGEN OR Kg i 5.0 ( 50) 5.0 (10-50)* 10 (10-50)*
TOTAL GAS CONTAINMENT GAMMA uCi/CC FACTOR OF TWO FACTOR OF TWO FACTOR OF TWO FACTOR OF TWO ATHOSPHERE SPECTRUM CONTAINMENT HYDROGEN PERCENT -i1 NA 1.5+ 11.5+
ATHOSPHERE
@ ORIGINAL ANALYSIS ACCURACIES WERE PROVIDED IN REFERENCE (e) 0F ATTACHMENT 4. .
- NRC (FRANK WITT) PRESENTED ANALYTICAL PROCEDURE ACCURACIES AT THE AMERICAN NULCEAR SOCIETY 1983 MEETING (REFERENCE (i) 0F ATTACHMENTS 4).
ACCURACIES ARE EXPRESSED AS 1-STANDARD DEVIATION (68% CONFIDENCE INTERVAL) UNCERTAINTY ESTIMATE.
" THE CONTAINMENT SUMP IS SAMPLED VIA THE RESIDUAL HEAT REMOVAL SYSTEM FOLLOWING A LOSS OF COOLANT ACCIDENT.
+ DETERMINE BY READING CONTAINMENT HYDROGEN ANALYZERS.
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ATTACHMENT l4 -
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! REFERENCE DOCUMENTS !
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-(a).' Letter from L. M. Mills, Manager of:TVA Nuclear. Licensing,' to J
.E. Adensam,. Chief'of NRC Licensing Branch No. 4, dated I
November 23, 1983.
-(b) Letter from L. M. Mills, Manager of TVA' Nuclear Licensing, to ;{
E. Adensam, Chief of NRC Licensing Branch.No. 4 ' dated i December 21,'1983. .$
(c) Letter from L. M. Mills, Manager.of-TVA Nuclear Licensing, to- >
E. Adensam, Chief of NRC Licensing Branch No. 4, dated January 9, 1984.
(d) Letter from L. M. Mills, Manager of TVA Nuclear Licensing, to -]
E. Adensam, Chief of NRC Licensing Branch No. 4, dated January 10, 1984. .
+ (e) Letter from L.-M. Mills, Manager of TVA Nuclear Licensing, to-E. Aden' sam, Chief of NRC Licensing' Branch No. 4, dated March 23, 1984. i (f) NUREG-0737,'" Clarification of IMI Action Plan Requirements" ,
(Item II.B.3, "Postaccident Sampling Capability" Provided).- :
.1 (g) Regulatory Guide 1.97, Revision 2, " Instrumentation for Light-Water - ;
Cooled Nuclear Power. Plants to Assess Plant and' Environs Conditions' !
-During and'Fo11owing an Accident" (Postaccident Sampling Requirements- l Provided). !
(h) Letter From NRC to TVA dated August 22, 1991, " Instrumentation to Follow the Course of an Accident (R.G. 1.97).(MPA A-017)-(TAC i Nos.'51133/51134) - Sequoyah Nuclear Plant, Units 1 and 2."
l (i) American Nuclear Society 1983 Meeting Presentation by Frank Witt,
" Requirements for Postaccident Sampling and Analysis," (Postaccident Range and Accuracy Requirements Provided). ,
(j) Letter SECY-93-087 from James M. Taylor, NRC Executive Director for-Operations to NRC Conunissioners, dated April 2",:.1993, " Policy, Technical, and Licensing Issues Pertaining to. Evolutionary.and-Advanced Light-Water Reactor (ALWR)' Designs'.' (Section II.I, R "Postaccident Sampling System," Provided).- I (k) NUMARC/NESP-007, Revision 2,'" Methodology for Development of .
Emergency' Action Levels," dated January 1992 (Criteria for Potential 'I Loss of Fuel Clad Barrier Provided).
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(1) : Regulatory Guide 1.101. Revision 3,' " Emergency Planning and-
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-Preparedness for Nuclear Power. Reactors," dated. August 1992.
(m). Letter.from James E.LRichardson, Director, Division'of..
Engineering, Office of Nuclear-Reactor Regulation to' John J.
Hutchingson, Florida Power and Light, dated April 12, 1993.
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a ,, ' ENCLOSURE.5'
, . . s FINAL SAFETY EVALUATION-REPORT. -
CHANGES' ASSOCIATED WITH SEQUOYAH
- NUCLEAR PIANTS' REVISED POSTACCIDENT '
-SAMPLING PROGRAM (TVA-SQN-TS-94-15).
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TABLE 7.5-2 (Sheet 5) -
TABLE OF VARIABLES FOR POSTACCIDENT MONITORiuG variable Type / Minlaus Mininnsa Redundancy Description ME C Ranee From Rarwe To Remired Notes Plant, Environs RAD-PORT Inst E3 1E-3 1E6 RADS /NR N/A (Photons)
Post Accident Sampling (PAS) -- -- -- --
(on Site Analysis) Depends on Sample See Below PAS, containment Air E3 See Notes GAMMA Spectrue N/A Isotopic Analysis
"'I, C r t:'__..: "'-
E! O 3^*- /A "2Crtrt
- i, Crt * -- !^'- E! O 3^* %/A Y,;-Cr*T* ~
PAS, Primary Coolant & Step 53 E3 50 6000 PPM N/A See Deviation No. 28 Boron Content PAS, Primary Coolant & Surp E3 . 0.1 20 PPM N/A Chloride Content PAS, Primary coolan !*- E3 10 2000 cc(STP)/kg N/A Dissolved H2 or Tote as See Deviation No. 22 I go I
nie .. _ _ _
c u_ _ __ ..._..__m_ n_
_ a_ _ e s_n_- u s.
n.____. _ = -_ m ._ _ ._ ..
s_
PAS, Primary Coolant .A Susp E3 See Notes N/A Isotopic Analysis GAMMA Spectrue PAS, Primary Coolant &_ Supp C3 E3 10 1ET uti/ml N/A Gross Activity PAS, Primary Coolant & Susp E3 1 13 Ph- N/A Ph Pressurizer Neater Status 02 0 800 AC Aaps N/A Pressurizer Level A1 81 0 1001 OF 3 Channets C1 D2 Actual Renee 210 to 474 Inches, Cytindrical Trend Recorder Required.
Portion
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.(f ):. . DEVIATION 22
, VARIABLE (111)
Postaccident Sampling. i!
DEVIATION FROM RG 1.97 GUIDANCE ,
i '
' The RG 1.97, Revision 2 (refer to Table 2, Type E~ variables),: recommends that ,
' primary coolant grab sample capability exists for hydrogen analysis.
s, ,
u JUSTIFICATION 4 -
i L- -
4QNie-poeta :idcat : T.pling f :ility ("ASF)-eveeently ::: inline escurr:nt of pr --ry eeelent hydreg:n by ge: ch: r tegr:phy.
4 32 hup :thed: t: inlin:
-nitering ese :::: :nded in Ite: II . " . 3 c f h"J"EC 0737 and' Tab 12 2 (Typ: :: -:
verf abler) in FC' 1.97; 5:=ver, SOM did' net ec-!t te
- hie espebility in sny ef T'!A' ' letter: referen::d in SQM': licence renditien. Further, m 3 incorrectly n:ted in th: :sfety eva'"-*4aa #a- 14a-a-- -- n d -n * ="rhae4*ing-snu'c nnermer4a.nr e. p11n8-.4 PAS).4ystem-operation-that-backup-methode-exist -
fa" ="=1ysing dissolved-hydrog'en in roaster-coolant crple:.
T cddeeee-these ;
i
-inconsisteneies, - TVA hee-- esued-a-condition adver:: t: qu Mty-eepe-et- ,
Q SM890A0?L Presen a part of the corrective-accion to 6Qre T402, SQN.
n' Nuclear Engineer g s conducted a design study.to define'the scope and
./ schedule.for modifying SQN's PASF.to provide hydrogen grab sampling .;
\ . ' -capabilities with a range of-10 to 2,000 cc standard temperature and pressure / kilogram (STP/kg). TVA considers this' range to be sufficient for i
estimating postaccident core deyggdatio, nf thg 7 rimary c quert: th:t =.C p::v-_: :=ter:: appr val of ::QN e tropesod- @devietion until plant ::dificati n: t: SOM*: PASF can be ca pleted
-(Cy:10 5- refueling :tege for both unite).
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2 -
. VARI BLE:(3) ,
ng Containment' Hydrogen Concentration.
, DEVIATION FROM RG 1.97 GUIDANCE e '
. The; range recounended 'in RG '1.97,, Revision ' 2,,is 0 ~ to .30 percent, whereas SQN -
1 --has provided instrumentation for this variable with'a: range of.0-.to 10-percent.
, JUSTIFICATION- m*
y ,
SQN has performed.an analysis.that shows~the worst-case hydrogen concentration-
, will be:less-than 8. percent with~the glow ~ plugs-(hydrogen. igniters).-
operating.- Thus, the instrumentation will be on. scale at any particular time..
-The hydrogen igniters' operat o 1 in t
_ tha c h room b (igator._ i t 'and larms. Additi:::lly, SQh": p::t :-id::: 2xplin;; ;y;;;;;
i o een re-ide di~erre indicatien fer this variable. concurs with SQ a Dnge pTcVilled of 0 o.10 percent. s a equa e. '
(
Reference:
iNRC-letter to S.'A.' White dated May. 11, 1987; " Emergency Response-Capability'- Conformance-to' Regulatory Guide 1.97, Rev. 2 [ TAC 51133/51134]") "p t
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"$d LDEVIATION7 $8
. VARIABLE (111): l J
P . .
.- Postaccident Sampling _i
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q DEVIATION FROM RG 1.97 GUIDANCE !
if . _
j RC 1.97, Revision recommends'.that-the' analysis' range for boron:contentuin-
~
2,.. :1
,/ the primary coolant and. sump be between 0 to 6,000 parts per million (ppm).. '
TVA recommends.that.the range be between 50-to'6,000. ppm.
c JUSTIFICATION For boron concentrations below 500 ppm, the tolerance for SQN's ?
Linstrumentation would be limited to plus or minus 50.. ppm. This tolerance band
^
. is considered by TVA to be acceptable for' ensuring that postaccident. shutdown '
margin is maintained. TVA's position is that the current range. capability.for'-
boron analysis (50 to 6,000 ppm) is sufficient.
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SQN-6
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~, Air flow is from areas of lower radioactivity potential to areas of greater radioactivity potential. All exhaust air is monitored for excessive radioactivity levels.
Fire dampers are used to prevent the spread of fire between the CDHEB and
- the waste package area of the Auxiliary Building.
9.4.9.2 System Description The CDWEB ECS is shown on Figures 9.4.2-1 and 9.4.9-1.
Air induced by the CDHEB supply fan from the waste package area supply duct is used for building ventilation. The ventilation air is supplied to areas of low radioactivity potential and migrates by naturally induced flow paths to progressively higher areas of contamination.
~
The CDHEB ventilation exhaust fan exhausts air from the area with highest contamination potential and directs it to the Fuel Handling Area Exhaust System where it is passed through a radiation monitoring station prior to its release to the atmosphere.
The CDHEB utilizes one speed ventilation fans. The fans are manually controlled and operate continuously.
Additionally, separate air-conditioning recirculation systems serve the (n\ potentially contaminated areas and the moderately contaminated areas.
'"J 9.4.9.3 Safety Evaluation No nuclear safety-related systems or components are located in the 2 Condensate Demineralizer Haste Evaporator Building. Therefore, a single failure within the EC System will not affect nuclear safety.
9.4.9.4 Inspection and Testing Requirements The CDHEB ECS will be tested initially to assure that design criteria have been met. Continued satisfactory operation will demonstrate the system capability.
9.4.10 Postaccident Sampling Ventilation System -
9.4.10.1 Design Basis The postaccident sampling facility environmental control system (PASFECS) provides heating, cooling, and ventilation during normal plant operations and training activities. In addition, heating, ventilation, and control of airborne radiological contamination is provided during postaccident acquisition and testing of samples. This is accompli %hed through l pressurization of the areas by the ventilation system which induces air from areas of lesser to areas of greater contamination potential. The
,3 system maintains temperatures within a range of 50*F to 104*F. The l6
/ ) PASFECS has redundant isolation capability in all ductwork which
- j 4
interfaces with the Auxiliary Building Gas Treatment System (ABGTS) or penetrates the Auxiliary Building Secondary Containment Enclosure (ABSCE).
9.4-41 0083F/COC4
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/f 9.4.10.2 System Description b The PASFECS is shown on the following figures:
9.4.10-1 (Flow Diagram 47W866-15), 9.4.10-2 (Logic Diagram 47W611-31-9),
2
)
and'9.4.10-3 (Control Otagram 47H610-31-9). The PASFECS consists of a ventilation subsystem (PASFVS), a heating and cooling subsystem (PASFHCS), and a radiological. gas treatment subsystem (PASFGTS).
(
9.4.10.2.1 PASFVS During normal plant operation,. ventilation air is supplied to the facility via the Auxillary Building general ventilation system and an auxiliary supply fan. Exhaust air is ducted directly to the Auxiliary Building general ventilation system.
During postaccident conditions or sampling operations, the normal supply
] and exhaust systems are isolated and ventilation air is taken directly 4
from the outside at a point on the roof of the unit I additional equipment building. Both the unit I and unit 2 systems share this common intake. A supply fan provides air to the sampling side of the facility in response to a differential pressure controller. Air is drawn from both the sample and valve gallery areas by an exhaust fan and routed to the exhaust duct downstream of the ABGTS air cleanup unit. The sampling. ,
n area is maintained at a positive pressure 2 0.12 inch WG with respect to
- e s atmosphere while the valve gallery is kept at a negative pressure of h 10.25 inch WG with respect to the sample side.
9.4.10.2.2 PASFHCS In the normal mode of operation, supply air taken from the Auxiliary '
Building general ventilation system has already been tempered and no j additional heating or cooling is required.
p In the postaccident mode, incoming air is preheated in response to a duct' mounted temperature switch. No cooling is provided in this mode. .
) However, the ventilation system will maintain the facility below 104*F
~
b with 97*F outside conditions.
i 9.4.10.2.3 PASFGTS .
The radiological gas treatment subsystem consists of one
$ HEPA/ charcoal-type air cleanup unit located just upstream of the exhaust fan. Air supplied to the facility during postaccident conditions or sampling operations is processed through the air cleanup unit prior to being discharged to the atmosphere. '
9.4.10.3 Safety Evaluation u L j The PASFECS is not a nuclear safety related system; however, the j
{ 4 isolation valves and duct which interface with the ABGTS and ABSCE are J 9.4-42 0083F/C0C4
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':1 ) .' designed to Category 1 standards. These valves are also backed by Class N._/ IE power. .All remaining portions.of the system are designed to Category ,
1(L) requirements.
9.4.10.4 Inspection and Testing Requirements-Prior to power operation tests were performed to assure that'the Post
'. Accident Sampling Facility Ventilation Subsystem performs as designed. .,
The Post Accident Sampling Facility Ventilation Subsystem will be 5-periodically inspected and tested in accordance with Plant Inservice Testing Procedures. Compo.nents will be subjectec to periodic inspections and tests in accordance with approved plant maintenance program procedures.
Air cleanup units are designed and tested per the requirements of NRC '
Regulatory Guide 1.140. Preoperational tests provided data for the initial balance of the system and verification of design flow rates.
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SQN-5 qqm 9.5.10 Postaccident Samoling 'FacilityJ k~. .
9.5.10.11 Desian Basis
- The posta Jden ampling facility'(PASF) is designed to safely obtain,-
- . transfer ant.yze. nd dispose of,-as necessary, samples of: reactor-
- coolant, co nt. sump water, and the containment atmosphere samples.
, Each reactor unit has its own respective PASF that will obtain the necessary samples following a loss of coolant accident (LOCA).
'9.5.10.2 Facili ties -
-The major components of the postaccident sampling system (PAS) are discussed in the following sections. swpt.m c 9.5.10.2.1 Reactor Coolant and Containment Sump System Each unit has a reactor coolant sampling system equipped with a closed cooling water heat exchanger to cool the sample as it-is' acquired by.the liquid sampling panel (LSP). Samples are taken from the reactor coolant hot legs and from the containment sump, when the RHR system.is l'n the recirculation mode of operation. 4 5
'9.5.10.2.2 One,,ical f nelvsi s Syste;;;
O The :::ple t:A:n fro;;; the !.Sa is re;.t;d te :he erdi,,, ch..i.;ce; er,eljii; pace! (CAM eher: th: f0!?ce h; :n:1y::: :r: p:rfer--d.
- 1. Hydre;:ncen:entr:ttenj:!;;::':hrrt:;r: phi l
- 2. Ch!cr!de cencentratten using i^^ exchin;e ch-a='t^araphy
- 3. Ioni cond;;tivity l
j T;;;;r:ter:
5
- 5. 01::Olved 0:yger concentrat!cn
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- 5. p" i I
- ni!y:! 2 three;5 pp'.s u u. 5 are perfor-ad
_.... ..a . . . , u m. by c^a"^ntiona!
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F -fer the 5n$iyEA$ hE$h5N T U hbY$'U11bT15'hA((5s e'quipM
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' with flush WATER '
nes [ ~~ ' dUscharg5Ti5D55F COIIector. Orain Tank which is drained into the tritiated drain collector. tank or the -
a containment sump.
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A
)I Containment Air Sampi a
ystem Acquisition of the-containment air samples is performed by the .
Radiological and Chemical . Technology (RCT) particulate, iodine, and_ gas !
. separation system and containment air sample panel (CASP), jointly. ,
These samples are subsequently transported to an onsite facility for isotopic analysis. Hydrogen levels in the containment atmosphere are determ ned4y--the._ronte'
,5 A rv1 PtJ N C AoD ent hydrogen monitors.
9.5.10Mi Analysis cArasivvie s a onstre Samples acquired in the PASF will be_trans r t ,the radicchemic=1 borat wher a not aer# creed in the PA.. be complete 6' PRoWmd
.rfe .wAust.s suPro a a 1sn so , w ,=n ci m .rs*uw t
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500 Section 9. 5.10. 2 . 2.-
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-9.5.10.3.2 "adlochem!ca1 Lah^ratory The radiochemica! 1:b0ratory ei!' be used to :n !y::e-the s =ples-take fro = the reestor-coolant. containment-sump Nh n P"" syst m is it ,
recJrculation edel -and_contain= v ant atmosphere. These analyses include,
,, but are not be limited to the following:
I \
(#/ 1. Isotop!c analy444-wt4Lbe-acc^=p!ished4y-obtaining-gamma-spectra--
4geraanius-detectorAof 14 quid-and-air-samp-1es using-es-tam 1shed procedures in the counting-room-locatebin-the-radiochesteal laborator-y.
- 2. Boren :nalysis-can be performed " the-e*4+t!ng p!:-t radio chemical t
laboratern - . _ _ - -_
9.5.1 . Design Evaluation .i The design life of all major components, equipment, and instrumentation l 1s 40 years. Items designed for postaccident service will be designed to remain 4 nctional in the expected postaccident environment. l 9.5.1 .) Tests and Inspections f The equipment located in the PASF will be tested and inspected to verify ,
equipment operability and availability.
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9.5-21 0084F/COC4 j
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-POSTACGIOENT REACTOR vuC^^lANT WATffMWt.ittE- ,
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z TABLE 9. 5. 10 - 1 M. i SEQUOYAH NUCLEAR PLANT REVISED POSTACCIDENT SAMPLING PROGRAM ,
r SAMPLING / ANALYSIS ANALYSIS SAMPLING /ANALYS1$ SAMPLING / ANALYSIS UNITS RANGE ++ ACCURACY ACCURACY RESPONSE TIME SAMPLE TYPE-SAMPLE POINT PARAMETER 1 5% (1000 6000)* 1 10% (500 6000)*- 8 HOURS G,# GRA8 SAMPLE RCS AND/OR CONT. 8ORON PPM $0 to 6000 SUMP ** 1 50 (50-1000)* 2 50 (50-500)*
FACTOR OF TWO FACTOR OF TWO 24 HOURS # GRAB SAMPLE -
RCS ANO/OR CONT. GAMMA uti/mL ISOTOPlc SUMP ** SPECTRUM - ANALYSIS 10 to 1E+7 FACTOR OF TWO FACTOR OF TWO 24 HOURS # DETERMINE 8Y-RCS AND/OR CONT. CROSS uCl/mL '
TOTALING GAMMA SUMP ** ACTIVITY ISOTOPIC ACTIVITIES 7-PPM 0.1 to 20 2 10% -(0.5-20)* 1 10% (0.5-20)* 24 HOURS (SAMPLING)# PROVISIONS ARE' RCS AND/OR CONT. CHLORIDE
% HOURS (ANALYSIS)# ESTABLISHED FOR OFF' -9 SUMP ** 1 0.05 (0.1 0.5)* 1 0.05 (0.1-0.5)* '
SITE ANALYS!$
24 HOURS # ' GRAB SAMPLE RCS DISSOLVED CC(STP) 10 to 2000 1 20% (50-2000)* + 20% (50-2000)*
HYDROGEN OR Kg 1 5.0 (10-50)* 15 (10-50)*
TOTAL GAS FACTOR OF TWO FACTOR OF TWO 24 HOURS # GRAB SAMPLE CONTAINMEhT GAMMA uCl/CC- ISOTOPIC '
SPECTRUM ANALYSIS P
ATMOSPHERE NOT APPLICABLE DETERMINE BY' CONTAINMENT HYDROGEN PERCENT 0 to 10 1 1.5 11.5 READING CONTAINMENT ATMOSPHERE HYDROGEN ANALYZERS
- THE CONTAINMENT SUMP IS SAMPLED VIA THE RESICUAL HEAT REMOVAL SYSTEM FOLLOWING A LOSS-OF-COOLANT ACCIDENT.
++ THE SAMPLING / ANALYSIS RANGES HAVE BEEN APPROVED AS PART OF THE SEQUOTAN NUCLEAR PLANT REGULATORY GUIDE 1.97 FINALIZED PROG
- ACCURACIES ARE EXPRESSED AS 1 STANDARD DEVIATION (68% CONFIDENCE INTERVAL) UNCERTAINTY ESTIMATES.
3 IF POTENTIAL CORE DAMAGE IS NOT INDICATED FOLLOWING AN ACCIDENT, THE RCS WILL BE SAMPLED AND BORON AdALYZED TO VERIFY SMUTDOWN MARGIN AS SOON AS
.POSSIBLE IN RESPONSE TO EMERGENCY PROCEDURES.
- FOLLOWING A POTENTIAL CORE DAMAGE ACCIDENT, SAMPLING / ANALYSIS WILL BE COMPLETED WITHIN THE STATED RESPONSE TIMES AFTER REGAINING STABLE CORE CONDITIONS.
- . . _ _ .-. . . - - .~~-a...- , , ,-, -- .-. - . . . - . - - - - - . - - . - . .