ML032950149
| ML032950149 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry, Watts Bar, Sequoyah  |
| Issue date: | 04/25/2002 |
| From: | Tennessee Valley Authority |
| To: | Office of Nuclear Reactor Regulation |
| Byrdsong A T | |
| References | |
| +adjud/ruledam200506, -RFPFR, 50-259-CIVP, 50-260-CIVP, 50-296-CIVP, 50-327-CIVP, 50-328-CIVP, 50-390-CIVP, ASLBP 01-791-01-CIVP, RAS 5907, TVA-Licensee-93 | |
| Download: ML032950149 (188) | |
Text
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Problem Resolution ',heet Closeout:
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1P1 Open Corrective Action entered into tracking system.
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SEQUOYAH NUCLEAR PLANT -
CLOSURE OF INCIDENT INVESTIGATION (II) JZ rsq 1 0 All action closure documentation has been reviewed and the actions taken were ade.ji;;t;:.
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?ACIAGE
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Date Received
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Has training been conducted in the area that contributed to the event?
Yes No Was a train task associated with this event?
Yes E /
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EXECUTIVE 8UMMARY
) 8 November 6, 1992 it was discovered that certain Technical
-Specification Gaseous Radiation Monitors could have had their setpoints calculated in a non-conservative manner. The subject Radiation Monitor setpoints do not account for the system design which has the gas sample chamber upstream of the sample pump, thus creating a vacuum in the detector gas chamber.
This pressure dfference requires a correction factor to be applied in order to calculate the setpoint.
Applying the correction factor to several Radiation Monitors indicates the actual setpoints could have been greater than the Technical Specification Allowable Setpoints specified in TS 3.3.3.1.
A related and concurrent event identified in the Problem Evaluation Report SQP900281PER indicated non-conservative vendor calibratior data had been used in the calculation of these same Radiation Itor setpoints.
This, when combined with the failure to a. unt for detector pressure in the setpoint calculation, c.
cunded the error.
Since the action.
TS 3.3.3.1 was not applied, it is concluded that SQN was in a cL ision prohibited by Technical Specification.
Upon covery plant personnel took readings at the Radiation Monitors..
ensure they remained within their whnical Specification i l. %le Setpoints. The Instrument KA function alarms were ti.
calibrated lower on the subject Radiation Monitors to ensure Technical Specification compliance.
The root cause of this event is inadequate engineering control of Radiation Monitor setpoints and setnoint methodology.
Contributing causes were fa ure o. anagement to ensure that adequate interdepartmental
.ommunic Ins existed and responsibilities for the o'erall Racism.. ',n Monitoring System were not well defined or understood.
FIOOOJ-c; 1
DESCRIPTION OF THE EVENT
\\
- i. Initial Conditions On November 6, 1992 it was discovered that certain Technical Specification (TS) Radiation Monitors (RM) could have had there setpoints calculated in a non-conservative manner.
The subject RM setpoints do not account for the system design which has the gas sample chamber upstream of the sample pump, thus creating a vacuum.
The original gaseous calibration by the manufacturer was performed at atmospheric conditions.
The operating pressure difference requires a correction factor to be applied in order to calculate the correct setpo-'t for the RM's.
Units 1 2
re operating at approximately 100% rated thermal p. -
The Noble Gas alarm/CRI setpoints of the Containment atmosphere monitors RM-90-106 and 112 and the Containment P -
monitors RM-90-130 and 131 were at 75%
of the TechniL Spetcification (TS) limit.
The high vacuum malfunct.
arm for the subject RMs was set at 12 +/-
inch of Me.
ry (IN Hg) below atmospheric pressure.
There w no -ressure correction applied to
) the Noble gas monito Go.
'ent Vent Isolation (CVI) setpoint.
Assuming th.
ic... m could be as high as 13 IN Hg without requiring operator action or initiating a CVI, the setpoint could be 1131 -f the TS limits of TS 3.3.3.1.
B.
Sequence of Events The Chemistry Sec'.on procedure TI-.6, Radiation Monitoring, is tha procedure for controlling and calculating the aarm vatpoints for Ms. The General Atomics (GA) 1974 Calitz :ion Report provided the detector sensitiv.ty val es for the initial developement of this procedure.
Thz L:.tial. procedure, TI-18, set the CVI setpoint at lot of he TS limit for both the particulate and 'Ionle gas detectors for containment atmosphere monitors R 10-106 & 112. The containment purge Noble gas mnitc RM-90-130 & 131 CVI setpoint was set at 10% of the TS 1 -nit.
This was sufficient margin to compensate for gas etector chamber pressure errors and for the use of n conservative gas detector GA calibration senstivi :..es.
kInitially, TI-3 set i:he Main Control Room (MCR)
Noble F I t:GC..
gas monitors 1.M-90-125 126, setpoint at 350 counts per minute (cpm).
The TS limit is 400 cpm.
When corrected for the highest possible vacuum prior to alarming, the CPT setpoint could have been 529 cpm.
Kjprii of 1979 GA issued a revision to the Calibration Report.
This report reduced the sensitivities for the gas detector making the earlier sensitivities non-conservative.
The report was issued to the Engineering Design organization in Knoxville.
The plant instrument engineer eventually received a copy of the report, but the Chemistry Section never received a copy.
On April 28, 1982 revision 10 to TI-18 raised the Noble gas CVI setpoint on R-90-106, 112, 130, and 131 to 40%
of the TS limit. When this setpoint is corrected for non-conserva tIv' detector sensitivities and chamber pressure,.th. ".T could have occurred at 93% of the TS limit.
The NRC issued t* -e 82-49, Correction For Conditions For Air And Gas Mc D.'ng.
This notice identified that pressure correctio.
- ne.cessary to account for sample deviation from Standa.
AT ature and Pressure (STP) conditions in gas dete, rotometers.
The notice was entered into the Nuck
-?erience Review (ER) process and sent to all appl skate organizations. The Ir "rument Maintenance Group sp -3ed that the
-\\
ration of the EM rotomett, -.aaople flow was corrected
-r pressure. The Chemistry SecLion responded that analyst take into account pres sure differential within monitors when collecting parti ulate and charcoal samples.
No one addressed th
-fects of pressure on the RM Noble gas detectors.
No c-l active a tion was identified.
In 1986 and 1987 a Surveillar e Instruction Ev.rification and validation program was e ablished to ensure the complete accuracy and validit4 f all plant procedures used to comply with the TS.
Wa ts Bar Plant personnel were acquired to perform this r view for Sequoyah.
The April 1979 revision to the GA Calibration Report was identified.
The revised repo'-t was included in the references section of TI-18, t the revised detector sensitivity values were not e ered into the procedure.
On April 16, 1990, nstructic Change ICF-90-217 changed the Noble gas CVI setpoint tc.. M-90-106 and 112 to 70% of the TS limit and removed thr tpoints to RM-90-130 and 131 from TI-1.
ICF Be td the CVI setpoints for IM---130 and 131 to S Containment (Upper, Lower) u,
-and ra he gas CVI setpoint to 70% of 3
F OCU
the TS.
On April 26, 1992 the setpoints in the above monitors were raised to the values allowed by the procedures.
When the setpoints are corrected for the non-conservative detector sensitivities and chamber pressure, the alarm could have occurred at 143% of the TS
.KJ limit.
On June 12, 1990 A Problem Evaluation Report, SQP900281PER was issued by the Chemistry Section identifying that the April 1979 GA Calibration Report had not been implemented in plant procedures.
An operability evaluation determined that the current Noble gas CVI setpoints for the RM-90-106, 112, 130, and 131 were at 95% of the TS limit.
This operability evaluation was based on the current containment atmospheric ic'ope mix, best estimate isotope sensitivities obtained from the 1979 GA Calibration Report, and current CVI setpoints of 70% of the TS limit.
When the CVI setpoint is corrected for detector chamber vacuum, the CVI could have occurred at 143% of the TS limit.
On Decembe 27, 1990 SI-410.2 revision 12 and TI-18 revision ere issued.
These procedure revisions included the 'etector sensitivities from the April 1979 GA Calibration eport and raised the Noble gas CVI setpoint to 75 f the TS limit.
The sensitivities were about 5% less t n t
. used in the SQP900281PER operability evalL 1).
because a more conservative t) yapproach was used interpreting the GA calibration A->' data.
However, wher.
e able gas setpoint were further corrected for chamber
%ssure, the CVI could have occurred as high as 113% albove the TS limit.
Revision 24 to TI-18 also It er e CRI setpoint on the Main control Room Noble gas..
i
-s RM-90-125 and 126 to 253 cpm.
When this setpoirnt is c eacted for chamber pressure, the monitor could CRI high as 382 cpm.
his is less than the TS limit of 400 In.
C. Inmediate Corrective Actions Based on a table, "Correction I r Sample Chamber Pressure", provided by the vent 3r for the model RD-52 gas detector the System Engineer d termined that if the Noble gas detector chamber vacuum did not exceed 10 IN Hg below atmospheric pressure, there was sufficient margin in the CVI setpoint to accommodate the chamber vacuum error.
This option was chosen over rel ing the monitor CVI setpoint, because lowering the ( 'I setpoint would increase the chances of initiat.g a CVI.
The following 1\\_
actions were initiated:
FIOOOIUIa i
- 1. The vacuum in RM-90-106, 112, 130, 131, 125, and 126 were immediately verified by the System Engineer to be less than 10 IN Hg.
- 2. The MIG Group initiated a program of reading the vacuum on the above monitors twice per shift to verify X>
the vacuum to be less than 10 IN Hg.
- 3. The MIG procedures SI-302 and SI-302.2 were revised to lower the vacuum alarm seLpoi.% to 9 +/-
IN Hg.
- 4. The performance of SI-3n' and 302.2 with the revised vacuum alarm points was completed on November 14, 1992 and the twice per shift reading of the RM vacuum levels was suspended.
ANALYSIS OP TE BVEMT A. Evaluation of Plant Systems and Components The two problems identified in this incident investigation, non-conservative Noble gas detector sensi ivities and no correction for gas chamber pressure, are a licable to all GA model RD-32 Noble gas detectors.
the spe.-fic monitors are as follows:
1,2-RM-z-99 O-RM-90-101B I
1,2-RM-90-l06B UJ 1,2-RM-90-112B O-RM-90-118 1, 2-RM-90-1' 9 I
0-RM-90-125 I 0-RM-90-126 I 1,2-RM-90-330 1,2-RM-90-!31 0-RM-90-11,2B 0-RM-9C-205 0-RM-90-106 I TS monit.ors The MCR Noble gas monitors RM-90-125 and 126 were initially calibrated with the CRI setpoint at 350 cpm.
When this value is corrected for gas chamber pressure, the CRI could occurred at 529 cpm.
TS CRI setpoint is 400 cpm.
This out of TS condition existed from initial licensing on unit I to December 27, 1990 when the CRI setpoint was reduced to 253 cpm (382 cpm when corrected for chamber pressure).
the cont;
-ment atmosphere monitors, R-90-+/-O6 and 112
_.X>I setp ts were initially at 10% of the TS limit.
The 5
FI 010
containment purge Noble gas monitors RM-90-130 and 131 CVI setpoints were also initially at 10% of the TS limit.
Though the Noble gas CVI setpoints were raised to 40% of
'he TS limit in 1982, sufficient margin still existed to l
commodate the use of non-conservative detector ensitivity values and applying no correction for gas chamber pressure.
On April 26, 1990 the CVI setpoint for these detectors were raised to 70% of the TS limit.
Late in 1990 the CVI setpoint was raised to 75% of the TS limit.
At this point sufficient margin no longer existed to bound the above errors.
These monitors remained outside the TS limits of TS 3.3.3.1 until November 6, 1992, when the gas chamber vacuum was verified manually to be less than 10 IN Hg below atmosphere. The vacuum was contir.ue-to be verified less than 10 IN Hg twice per shift unti t..
high vacuum alarm setpoint was changed to 9 +/-
IN Though the other monitors shown in the above list could also have non-c servative alarm setpoints, they are not TS monitors.
Ti.
monitor effluent stream, but have never been use t antify Offsite Dose Calculation Manual (ODCM) efflu ases.
All ODCM effluent releases are quanti_
s'amples analyzed in the Chemistry Laboratory test equipru i The Particulate and :odir.-.
nitnrs were calibrated with solid source as pposed to.- gas source for the gas yU,.litors.
Also, i thrcugh ch-ircoal and particulate filters is end alvays has been adjusted to Standard Cubic.-et Pr n te (SCFY).
The CVI setpoint for the RM-90-106, 11k, 1so, a..i 131 pArticulate monitors has always been maintained at less than or equal to 40%
of the TS limit.
- Thus, ufficient margin has always existed.
There is an Eberline Nble gay. detectors on the Condenser exhaust. This detector < not corrected for chamber pressure.
The Eer'ine l&":e.tor operates with a positive pressure in the chamber.
T' :efore, the RM output would be conservative.
There is a Sorrento No}le gas detector on the shield building exhaust. Thee etectors are digital detectors that have internal colren3ation for the detector chamber pressure.
INSTRUMENT INACCU)RACIES Per Memorandum "S"TE L tNSING POSITION ON INSTRUMENT T
7CURACIES1' (R' 1S S1C L 3878), SQN Licensing position L,
e sidificaoce of ument naccuracies in F I 00 0.2 2C
selecting Technical Specification (TS) surveillance acceptance criteria can be based on quantitative evaluation of the margin available in each analysis.
'Determinations could then be made as to whether or not Q
ufficient margin already exists in the safety limit determined by analysis to bound instrument inaccuracies.
0-RE-90-125 & 126 Per GDC 19, the exposure limit for control room personnel must not be in excess of 5 rem whole body, or it. equivalent to any part of the body, for the durat.:.
of the accident.
The 400 cpm TS limit is equiva.:
t o 1.76E-5 Ci/cc. Using Xenon-133 the dose rag to control room personnel would be 0.614 mrem/hr o rem for 100 days.
1.2-RE-90-106. 112. 130 & 131 Per 10CFR20, no.
ividual shall receive a dose to the whole body in -y riod of one calendar year in excess of 0.5 rem.
- ie 5 Radiation Monitoring setpoint is based if v
- tinuously Purged 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> a day/365 days a jar s, would not exceed the OCFR20 limit.
However, T (3.6.1.9) o-'y allows each unit to Purge for :--
hours per ye. -.
.nerefore, there is significant ma in in thn se.., int to bound the instrumei. inaccuracies.
This demonstrates that there
. sufficient margin in the development of the TS limit t encompass instrument inaccuracies.
B. Evaluation of ersDnne.
erfo.iance The evaluation of the NRC IE N
- .ce 82-49 was evaluated in part by several organizati
'the Instrument Maintenance Group evaluated ct of vacuum on rotometer indications.
They concl;.!d they already corrected rotometer 1ding for va:
tin to obtain SCFM values. The Chemistt, eortion eva.
ted the effects of vacuum on the char-al,
.rticulat : and gas samples obtained for anal' in the lab.
Though, TI-18 states that the Chemistr ection is respoisible for the Alarm setpoints for all Aft, they did no' evaluate the effects of vacuum on the RH gas detectors.
Tn the 1986-1987 time frame, a prgrawt1 was instituted to Fify and validate all procedure vused "o ensure TS K.Mipliance.
The requirements of h
"o-ram were that FI 005. '21 I
all methods, techniques, calculations and requirements were to be traced to there source document.
The engineers who evaluated TI-18 found the revised GA Calibration Report issued in April 1979, but did not YU) include the revised sensitivities into the procedure.
There is no indication that the NRC IE Notice was included in the review.
The procedure was not revised to include as pressure correction allowance in determining gas detector setpoints.
On April 16, 1990 TI-18 and SI-410.2 were revised to raise the Noble gas CVI setpoint on RM-90-106, 112, 130, and 131 to 70% of the TS limit.
The procedure review and revision process and the 10CFR50.59 process failed to identify the needed gas detector pressure correction and the use of old non-conservative detector sensitivity values.
Though there were other procedure changes that also failed to identify these deficiencies, the sign i.cance of the April 1990 changes are they allowed the p
-t to exceed the TS limits of TS 3.3.3.1.
C. Safety C, equences and Implications As stated abovs, 1 & 2-RM-90-106 and 112,
-RM-90-125 and 126 and 1,2-RM '0-130 and 131 were out of calibration such that the.r spr~ctive TECH SPEC dose limits during containment pulrgiA.: operation could have been exceeded.
t g However, due to the eactor coolant gross activities X_> during this time per..
and shield building vent monitors, there is a hi ' level of confidence that the TECH SPEC dose limit, wo..
have been exceeded in the event of a design basis a
- ies, r operational transients during conti inmE.
pu.. ing.
Justification to support this position :s P--
vi%>d below.
EVENTS EVALUATED The reactor coolant tctivitie d ring the period of time the radiation monitor etpoint-were out of specification will have a big impac: on te radiation dose to the environment.
Therefore, tis e luation will analyze the impacts of a Large break I CA, S 1 Break LOCA, and RCS Leakage, during containme purg.
activities with the subject radiation monitor out ot Calibration.
In addition to these evens ne Fuel Handling Accident inside containm t w valuatec.
These conditions constitute the ost
- r.
Tting events for the subject radiation monitors be ng incorrectly setup.
q, LARGE BREAK LOCA 8
FA :.)Oi2.
get
In the event of a LBLOCA, a safety injection signal would be generated 2.7 seconds Reference FSAR Table 15.4.1-7.
This signal will initiate a containment isolation signal sulting in the activation of the Main Control Room rgency Ventilation System (MCREVS) and the closure of the containment purge isolation valves.
SI-166 documents the maximum allowable stroke time for the containment purge isolation valves (. & 2-FCV-30-7 through 10, 14 through 20, 37, 40, 50 through 53, 56 through 59) as 4 seconds.
FSAR Table 15.4.1-la indicates the earliest nuclear fuel rod burst will not occur until 50.7 seconds.
Therefore, only normal reactor coolant activities, not design basis activities, would be released into containment at the time of containment isolation.
It is noted in this scenario that MCREVS activation and containment isolation would have been achieved successfully without the use of RM-90-106, 107, 125, 126, 130, and 31.
It is concluded that this scenario would not have resulted in an unanalyzed conditions in the LBLOCA analysis provided in FSAR 15.4.1 or the Environmental Consequences of a Postulated LOCA provided in FSAR 15.5.3.
SMALL BREAK LOCA In the eveit of a SBLOCA, a safety -jection signal would I' generated 58 seconds for a 2 inc 3reak Reference FSAR le 15.3.1-1.
As mentioned above, this signal would Yu1ult in the activation of the MCREVS and the initiation
-of a containment isolation signal without the use of RN-90-106, 112, 125, 126, 130, and '"
Also in this scenario, only normal reactor coolant ctivities would be released into containment at the tine of containment isolation; see FSAR Table 15.3.1-1.
is concluded that this scenario would not have resulted in an unanalyzed conditions in the SBLOCA analysis provided in FSAR 15.3 or the Environmental Consequences of Postulated LOCA provided in FSAR 15.5.3.
CALCULATED RCS LEAKAGES SQNAPS3-063, Offsite and Control Roo X Operator Dose Due to a Small Line Break LO-A, calculated radiation doses at the site boundary, low population zone, and the control room operator in the event of a 4 inc.:h small break LOCA.
"For simplicity this calculation assumed the leakage from containment went directly to the envi:onment for the entire 30 day time period (no EGTS).
This leads to very con-servative results as there is no hld up, dilution, d
sition, and filtration of the radioisotopes
->,.oidered as they are released from ontainment.
It was 9
F 0 i, 4
also assumed that there was no failed fuel" resulting from the SBLOCA.
The source terms used in this analysis were taken from ANSI/ANS-18.l/1984, Radioactive Source "erm For Normal Operation Of Light Water Reactors, which y
Here used to determine the expected reactor coolant ctivities.
It is noted this methodology has been incorporate into SQN design basis calculation SQNAPS3-047, Reactor coolant Activities In Accordance With ANSI/ANS-18.1/1984.
Clearly, these assumptions bound the scenarios in question including RCS LEAKAGE during plant operations.
The results of this analysis are provided below in REM.
Type of Control Room 30-Day Low 2-Hr Site Radiation Operator Dose Population Zone Boundary Gamma 7.19E-05 1.31E-02 1.lOE-02 BETA 3.92E-04 3.68E-04 3.03E-03 Inhalation 5.43E-02 1.58E-01 1.31E+00 The control room operators will receive doses much less than the OCFR50
'DC 19 criteria of 5 rem whole body, 30 rem beta, and 30 i thyroid.
The calculated 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> Site Boundary and the 3 ay Low Population Zone doses are well within the NRC CFR50 GDC-19 and 10CFR100 limits.
ACTUAL RCS ACTIVITIES g~nus far in this evaluat..
it has been determined only normal reactor coolant activit's would have been released into containment up to the time the safety injection signal would resuelt in the activation of the MCREVS and the initiation uf a cnt nment isolation signal independently of the subject
- iation monitors.
Therefore, the radioisotope; assc-ted with only normal reactor coolant would be mi ~Ad ir.
containment purge effluent.
Subsequently a review {.
' actual RCS source terms continuously being circula-ed in the RCS primary coolant and interconnecting syst ms during routine plant operations was warranted This task required a review of SI-54, Reactor Coolant E-Bar Dettr Lnation data to evaluate SQN RCS actua gamma acti.v y for normal plant operations over the past 12 years.
This data was compared to the gamma activity prof-ded in ANSI/ANS-18.1/1984. The attached
- aphs, Figures 1 and 2, reveal SQN Units 1 and 2 actual gamma activities are bounded by the expected primary co l;nt activities in ANSI/ANS-18.1/1984.
This trending indicates the plant has operated within 33% of the ANSI fANS-18.1/1984 RCS
- aamma activities.
KILED FUEL REPORTS Fl O;0i 2 Itl 4
Since SQNAPS30047 assumes the dose rates from the plants components were based on an assumed 0.125% failed nuclear fuel per reactor core, it was prudent to review
-- d compare the SQN nuclear fuel performance report t
Llowing each refueling to the calculated values of
_-Wiled fuel.
Given below are the actual numbers of failed fuel obtained from SQN Fuel Performance Summary Reports CORE RELOAD UNIT UNIT 2 FAILED FAILED FUEL FUEL CYCLE 1 5
4 CYCLE 2 5
4 CYCLE 3 2
6 CYCL 4
6 7
CYCLE 6
1 The above d a show a maximum of 7 failed fuel rods is bounding for age to date.
SQN FSAR Table 4.1-1 indicates the reactor cox contains a total of 50,952 fuel rods.
Therefore, SQNs
-ual worse case failed fuel ratio is 001374 which i.
ounded by a factor of 9 by the Kjamed 0.125% faile fuel rods.
This trending indicates
..he plant has operat. -ith a reactor core better than expected by the industr1 standards.
FUEL HANDLING ACCIDENT In the event of a FHA iside co,,*ainment during containment purging operations, t radioisotopes from the accident would be exhausted through the containment purge ducts.
The containment purge ducts houses the Shield Building Vent Ratiation Monitors located down stream of the Containment Purge Radiation Monitors.
These monitors do not provide a safety control function however, they do provide an alarm and indication in the main control room at 10' (31100 microCi/sec) of the ODCH or TECH SPEC value base_ on Xe-133, reference TI-18, Radiation Monitoring.
A mentioned earlier, the Containment Purge !'on-tors were incorrectly calibrated at 75% of this value.
T.erefore, the operator would have been alerted to take the correct actions to mitigate the event.
Furthermore, during nuclear fuel movement, TECH SPEC 3/4.9.5 requires direct
. munications be maintained b
ien the main control room nd the operators at the
.'r eiing station.
This reqLirement is administratively 11 A,
aA:
controlled via FHI-7, Refueling Operation.
FSAR 15.5.6, Environmental Consequences of a Postulated Fuel Handling Accident was performed in accordance with REGUIDE 1.25.
One the assumptions provided in REGUIDE K_/
1.25 requires all activity is assumed to be released over a two hour time period.
The results of this analysis indicates the doses are less than the lCFR100 values of 300 rem to the thyroid and 25 rem to the whole body.
Clearly the assumption allowing a release up to two hours bounds the requirement of direct communications be maintained between the main control room and the operators at the refueling station.
Therefore, it is concluded that this scenario would not have resulted in an unar. lyzed conditions in FSAR 15.5.6.
CONCLUSIOA.
Given the ex ting hardware and administrative controls that are in p e
e OCFR100 exposure limits would not have been ecee.-d ie event of a Large Break LOCA, Small Break LOCA A.
ikage, or Fuel Handling Accident inside containment ua. g containment purging activities with the subject raLiat.L n monitor's setpoint out of calibration.
These conli.'ons constitute the most limiting events for he ubject radiation monitor's U) setpoint being out of calibration.
II. EXTENT OF CONDITION The conditions of old noi-conser-fati, gas detector sensitivities and the aences o a gaL etector pressure correction is applicable to a.l C iodel RD-32 gas detectors.
The Eberline condense vacuum pump exhaust gas detectors do not compensate for g;.s chamber pressure.
However, they operate a a positive pressure which is conservative.
The Sorrento mode. RD-52 gas detectors are digital and internally compensa :e for detector chamber pressure. The intei-ial compenstion for gas chamber pressure in the mode. RD-52 is z. constant based on a value derived from a test.
PREVIOUS BSILAR ZVE14'TS A review of the Licen.ee Event P. alert LER), Nuclear Experience Review (NER), and th T icking and Reporting of sn Items (TROI) data bases we e c n4
'-ed to identify any vious similar events.
12 F 3 Io: 0{ :
The key words used for the NER search were RADIATION MONITOR and ETPOINT.
132 NER events were identified and none of these events were similar to this problem.
X )The same keywords were used in the LER data base search.
159 LER events were identified. These events were reviewed and determined not to be applicable to this subject.
The keywords used for the TROI search were all SYSTEM 90 documents.
422 TROI documents were identified.
One document was found to be applicable.
SQP900281PER was written to document the wrong sensitivities were being used to calculate RM setpoints.
The corrective action associated with this PER could not have prevented this event from occurring.
NRC E Notice 82-49, Correction For Sample Conditions for Air and Gas Monitoring, was identified by personnel involved in the investigation.
The Sequoyah response to this notice was inadequate.
No corrective action was identified.
Had this notice been properly evaluated, this event could have been prevented.
VI. ROOT CAUSE A. In 1979 a revision to the GA Calibration Report was K) issued.
A trace of the paper shows that this report was transmitted to the TVA Office of Engineering Design in Knoxville, Tennessee.
Though it can not be established how, the Sequoyah Instrument RM engineer received a copy of this revision.
The Chemistry Section never received a copy of the reviser report.
There did not exist a programmatic barrier to ensure that changes were reviewed by all appropriate organizations.
The root cause of was changes were not adequately communicated.
B. NRC IE Notice 82-49 was issued in 1982 that specifically deals with the effects of pressure on the response of gas detectors and the effects of pressure on the indication of rotometers.
The Sequoyah response was inadequate.
This is deemed an inappropriate action, because an adequate evaluation of this notice would have prevented this event.
The notice was evaluated in part by various plant organizations. The Instrument Maintenance Group evaluated the effects of pressure on rotometers measurements, and determined plant procedures adequate.
The Chemistry Section evaluated the effects of pressure on the iodine, particulate, and gas samples and determined the plant procedures to be adequate.
The K) effects of vacuum on the RM gas detector was omitted from F I 0001.24 A
the evaluation.
No one address the notice from the big picture, only in part.
The root cause was there was no methods in place to ensure interdepartmental communications.
The SI validation and verification program failed to identified the need to correct the RM gas detectors for chamber pressure.
The review identified the revised GA Calibration Report issued in 179, but did not include the revised values in TI-18. The cause of this omission could not be determined.
The personnel from Watts Bar who performed this review are no longer available to interview.
However, it is believed that their mindset would be the same as the indset in Sequoyah Chemistry Department. This barrier failed because management method did not ensure ownership oL all aspects of the RMs. The Chemistry personnel focus on effluent quantif ication and ODCM methodology.
The R setpoints and set; Aint methodology were controlled by plant procedux and did not adequately address all necessary aspects.
root cause is lack of engineering control of RM setpoinL nd setpoint methodology.
D. In April of IS-inadequate review was performed on revisions to the pl;Iant procedures which controlled the CRI setpoints for Ph's.
This action is determined inappropriate becau t'-e consequences allow the plant to violate the TS limits t TS 3.3.3.1.
The reasons and
\\_/ causes are the same as
'o. for the inadequate SI verification and val.'da.. n
.ons.
I. CORRECTIVE CTIONS A. There exist today an ER prcgram at will ensure all revised vendor information is revsi i for applicability by all appropriate organizations.
T.
, There is no action needed to corre' root cause A.
B. The action for root c ses B,, and D are as follows:
- 1. Define responsibilities for l aspects of RMs.
This action is complete.
An agreement on responsibilities is attached
- 2. Document responsibilities determined in action B. in appropriate pant documents.
SQO/CEM/WFJ; SQO/RAD/CEK; SQO r,!G/RDP;SQO/ICE/RXG; SQP/LMN/VAB Due 1/15/93
<-'3. Revise T-to address me';h.
o
-(counting for gas F MOVIEa.
detector chamber vacuum.
SQO/CEM/WFJ Due 12/11/92
- 4. Determine the bases for values of TS 3.3.3.1 and Table 2'-'
3.3-6.
Determine if TS changes are necessary.
SQP/LMN/VAB Due 1/29/93
- 5. Provide SSDs for RM with TS setpoints SQP/LEE/CRB Due 2/26/93
- 6. Revise the SSD for the R high vacuum malfunction alarm from a setpoint value of 12 +/- 1 IN Hg to 9 I 1 IN Hg.
Resp: SQP/LEE/CRB Cue 2/19/92
. actions ha.e been coordinated withthe responsible parties.
I.OTHER OBSERVATa " AND ACTIONS A. The NER database earch did not reveal the NRC IE Notice B2-49. The NER key:oi. indexes should be revised to ensure this notice will be picked up by searches for K.) Radiation Monitors.
Resp:
Licensing/Jim Smit'>
Complete F I ooo~*x l
l 2
i I
CI l
RESPONSIBILITIES FOR RADIATION ONITOR8 Maint -
K)
'Chem -
Rad Con-T/S -
NE -
Responsible for corrective and preventative maintenance, calibration of RMs Responsible for setpoints on effluent and process RMs (anything that readout in cpm or Ci/cc) and ODCH methodology including inaccuracies. They are considered the end user of the equipment and should oversee the daily performance including trending and results of RMs.
Responsible for RMs that read out in dose rate units.
They are considered the end user of these RMs and should ir. lude the oversight of RM performance, trending and re. Its.
- Proviae oversight for the RM system.
This includes maintenL e and calibration procedures as well as system health ar.
upgrade.
Provide the -Asign basis of the RM system. The are the owner of TS s,.
oints.
Jo F I 0C0 o
- i.
J.
. Holland, OPS 4D-SQN SEQUOYAH NUCLEAR PLANT -
INCIDENT INVESTIGATION (II)
ACTION ITEM EXTENSION/CLOSURE
Reference:
II Item No.
S 9 0
_0 II Action Item Sequence No.
O2.
I request that the referenced TROI action be
Ž closed as complete.
extended to Basis for extension/closure:
/~
For closure list upportin
^losure dpcumentation as required per SSP.12.9 (Attach Copy).
aB-M ~~~g r Y.3ac, /,! A/--
SIGATVREM(I/AC ON SUPERVIS;OR) /
bATE U
'/
I This extension does not impact nulcear safety or plant operability.
U Alternate Correct.ve Action.
I~~~~~~~~~~
EVENT MANAGER
/
DATE PLANT MANAGER I DATE OR DEPT MGR FOR CATEGORY 3 The responsible organization's and Plant Manager's approval (or Dept. Mgr's for Cate-qry 3) is required for all action data extensions and for closure when action taken is.
erence from the ap oved corrective action in the Final Event Report.
JE:PW902O5/3F I 003 PL090205/13i7 I
TENNESSEE VALLEY AUTHORITY SEQUOYAH NUCLEAR PLANT SITE STANDARD PRACTICE SSP-13.-i J
CONDUCT OF CHEMISTRY Revision 3 PREPARED/PROOFREAD BY: Mike Goodson DATE:
.SIGNATURL A t J-e Ca __-
RESPONSIBLE APPROVED BY:
REVISION DESCRIPTION:
DATE: /;
EFFECTIVE DATE: /1-l /.I-VERIFICATION DATE:
NIA I
i This revision modifies chemistry limits and sampling frequencies for various-plant systems.
Additional guidance for implementing and maintaining a data assessment program and recommendations for implementing and monitecring overall program responsibilities are provideC n this revision.
This revision will control systems 14 and 43 Chemistry annunciator/alarm setpoints to ensure compliance to current Chemistry standards.
DCNs HO09D2A and H08903A on Unit I and Unit 2 respectively removed annunciatorlalarm control from plant Instrument Tabulation drawings.
FI OO 3'
b 1 t SSP-13.1 STANDARD CONDUCT OF CHEMISTRY Rev. 3 PRACTICE Page 9 of 112 3.7 Bulk Chemical Specifications Strict criteria for system chemistry control at SQN requires that bulk chemicals and materials used In systems be controlled to prevent contaminant Introduction and ensure system operability.
Guidelines and requirements for bulk chemicals are delinerated in Chapter 10 of the Nuclear Power Chemistry Manual.
These requirements are based on regulatory criteria, plant system vendor specifications and industry best practices. The guidelines will be strictly adhered to for chemicals coming n contact with CSSC equipment.
The Chemistry Department shall verify the proper location for all site chemical and fuel ol deliveries and discharges. Control will maintained by locking all unloading and transfer valves.
C, rol of chemicals at SQN Is further delinerated in SSP-13.2 "Ci..
ical Traffic Control (CTC) Program".
3.8 Radloac
- e Effluents The Chentis, organization shall have a radioactive effluents monitoring prt a., as required by the site administrative Technical SpeL
'UW'ons, established n the Offsite Dose Calculation Hanu (C(2CM).
Corporate Chemlstry.
-e l-Ible for establishment and maintenanc' of a meteo.
program which complies with the requiremen. gven n 10.JR
- Q.
In accordanc.
,th 51'neloyah 1n..
Specification Administrative Section, the radioactive licuid efluent monitoring instrumentation channels shall be operable wth t. r alarm/trip setpoints set to ensure tit the limits of th ODCII a not exceeded.
Effluent rdlation monitor bacrgrounds d be monitored and maintained is low as possible to lnsure.W-ate effluent monitoring The setpoint for he effluent radiation monitor will routinely e set at a small fr t on above the expected response and will larm or automatically t rmirnate the release. The termination of the release will initiate an evaluation into changes In the release pathway, radioacd;vity levels, monitor backgrounds.
changes, etc.
Effluent release monitoring w also be evaluated for Regulatory G
e 1.21 compliance.
Repres
.1stive sampling verification,
'jent radiation nonitor res; i versus expected response, etc..
- ny anomalies will be evalu.
d d corrective action taken.
3.9 Ira. ictions FC co:
n1
- 3. H. Holland, OPS D-SQN SEQUOYAI NUCLEAR PLANT -
INCIDENT INVESTI(ATION (11)
ACTION ITEM EXTENSION/CLOSURE
Reference:
II Item II Act n
-j3sqzo( i
- Sequence
- o.
I request that the ref onced TROI action be
/
closed as complete.
/ / extended to Basis for extension/ciosure:
- :ctc'.
%Vks_.7A(3
\\;-Le":4
"-6
.. I'
(-A
^ 1 RA~z1A,
&-curei
't ' t I
-.- ^I I
N~~a~oW ?,.,,-,
4;w
-g
'i,~
-C,.
S For closure list supporting closure doc.,umentation' as required per SSP.12.9 (Attach Copy SIATUACTION S tU/
VISO
- !.I
-A TE iGNATURE ACTION SUPERVISOR ) I IIATE
-, tip I
-0 This extension does not impart nuclear rafety or plant operability.
/7 Alternate Corrective Action.
t DATE PLANT MANAGER The nnt managers approval is reqwir for a I Ion date extensions and for closnre wh ctlon taken s different fror, t a-c ed corrective action in the Final Event Oah_,y F1 0003.3 JHH: PMB
TENNESSEE VALLEY AUTHORITY SEQUOYAH NUCLEAR PLANT RADIOLOGICAL CONTROL INSTRUCTION RCI-5 RADIOLOGICAL CONTROL INSTRUMENTA7ION PROGRAM Revision 16 PREPARED/PROC-"EAD BY:
Steven R. Bradley DATI SIGNATURE: AL r,/4 RESPONSIBLE ORGANIZE i
-logical Control APPROVED BY:D hi_________
ATI E:
-/Z-5
- E:
EFFECTIVE DATE:
I-/si-1-1 1-1/5- -9.3 REVISION DESCRIPTION:
Revised tc ncorpora - mont'
-esponse checks for electronic dosimeter and
'nor editing changes.
F1000I-t
RCI-5 Page 5 Revision 16 5.6 Instrument Control (Continued) 5.6.3 RADCON portable Instrumentation available for use shall be In a well defined area separate from other instruments.
5.6.4 Individuals that use radiation protection equipment are required to adhere to the RADCON Instructions that dictate the use and handling of the equipment.
5.6.5 RADCON will monitor the availability of the plant general area continuous air monitors (CAM) and area radiation monitors (ARM) and will provide management oversight for these instruments.
QUALITY ASSURANCE PROVISIONS None.
Fl 000j.3(.
hldar
J. H. Holland, OPS 4D-SQN SEQUOYAH NUCLEAR PLANT -
INCIDENT INVESTIGATION (I)
ACTION TEM EXTEUSION/CLOSURE
Reference:
II Item No. 59Ao8,0 II Action Item Sequence No.
art I request that the referenced TROI action be /-/
closed as complete.
/
extended to Basis or extension (:ione
-r1Ce t
/ 5Xf&CA/s A/fv,A eAJHaA A.
4
.. ~ZCM,
m5.0G.=4( (1bA)Atl
,0
-e v..
1J7*":,-
4 X 14 K kos 43 B
A
_0- #:
-%//
J For closure list supporting closure documentation as required per SSP.12.9 (Attic n Copy) 4~~~.
0~~
g;2i2 e!O "
j
/
/
93 SIGNATURE (II/ACTPON SUPERVISOR) I DATE L
/ This extension does not impact nulcear safety or plant operability.
I Alternate Corrective Action.
Oti W-Itd
, 5.
L EVENT MANAGER
/
DATE 7
PLANT MANAGER
/ DATE OR DEPT MGR FOR CATEGORY 3 Tf-responsible organization's and Plant Manager's approval (or Dept. Mgr's for ory 3) is required for all action data extensions and for closure when dCLion taken v-dfference from the approved corrective action In the Final Event Report.
F I 0JH H,:
O-H: PM4B PL090205/1317
"43dart4t Shret lFrm o)N)' SIa) 4 4 1 lop I I
_I.
SITE SSP-6. 1 STANDARD CONDUCT OF MAINTENANCE Rev X-J PRACTICE Page 7 of 65
1.0 INTRODUCTION
Maintenance has a primary role in ensuring safe and reliable nuclear power facility electrical generation. To carry out this role, management endorses a strong maintenance philosophy based on a sound set of standards, values, convictions, and principles.
.'at philosophy is established and implemented by this Site
'dard Practice.
The E.
de, of this Site Standard Practice conform to INPO 85-03b ic.
nes for the Conduct of Maintenance at Nuclear Power Station.-.
1.1 Purpose This Site Standara.
ctice provides requirements, guidelines, and instructions for the duct of maintenance to ensure maintenance activities are conduct.
- n an effective, consistent manner in accordance wit the oerating ii."'nse, plant procedures, and applicable regulatory.equft ent,. Additionally, this Site Standard Practice prese.nts a.;.
dorses a professional code provided on Lopendix A e tena;... Professional Code.
1.2 Applicability This Site Standard Pr. t1.ce
.Lies to all personnel (TVA and contractor) involved i t the administration, planning, scheduling, supervision, and perfQ mance of maintenance ativities or functions at TVA's nuclear facili ties. The requirements of this Site Standard Practice also apply to maintenance activities performed by nonnuclear Power organiz tions on plant equipment.
2.0 REQuIeREnUrs 2.1 Maintenanet Organizathsa and A' ministration 2.1.1 Management Standards and Epecta ',ns In order for the ait enance Orgentvt tion to function effectively, the following management standare
.1 expectations are established for all Maintenance em loyees.
r, -I rif
. 1
'tJ
,jC.
'ir
-h.
I STANDARD CONDUCT OF MAINTENANCE Rev 0 PRACTICE Page of 65 l U
j.1 Management Standards and Expectations (Continued)
All Maintenance Employees A. STRIVE to achieve the highest level of tchnjcal competsnc! in his or her discipline--become the expert in their skill.
B. CONSIDER personal safety and the safety of others as being the utmost in the performance of any activity, AND COMPLY with site safety rules and the use of safety equipment at all times.
C. PAY attention to the minute details of each task performed, AND VERIFY each detail is correct, AND ENSL ' every activity Is done right the first time.
D. CuMMUNI, eectively:
- 1. ESCAATE problems.
- 2. REPOr.
eficit..
es quickly.
- 3. PROVITJi input to pi blem resolution.
- 4. ASK estions.
- 5. L'RI an 4 USE the communication systems such as conditions adverse to quality reports, work requests, and s gstion programs.
- 6. TALK t cheir supervise:.
It takes e employee's c trIbutlon to ensure success.
E.
SUPPORT a aam approach to act ities and the conduct of daily rout nes.
ItA I Yner their D oblem-- '
Is alwavs ours.
F 0 fO 04 iONP-1265)
-vt_.
STANDARD l
CONDUCT OF MAINTENAUCE Rev 0 PRACTICE Page 9 of 65 UJ-l Management Standards and Expectations (Continued)
All Maintenance Employees (Continued)
F. THINK problems and tasks through--be creative when deriving resolutions.
Application of new technologies and approaches is promoted.
G.
BE accountable for the quality and efficiency of his or her work, including technical correctness of the work, housekeeping, safety, and so forth.
B.
CONSIDER dose reduction in every task as low a reasonably achievable (ALARA),.
Is ire a better way to do the task at hand to minimize dose?
I.
CONSi.
shielding, flushing the system, use of glove bags, and so f h.
J.
RESPONSIBLE
- r.
- to work both mentally and physically fit to perform thel oJ.Ies.
K.)
The requirements of th, witness for Duty Program are delineated in SQA80.3.13 Fitnes.
- o. Duty Program.
K. MKINTAIR the plant to destin conditions, AND OBTAIN approval for wiy dtvi-...o. from design configuration by the appropriate design ontrol process such as a design change request and so forth. in
-ordance with the following implementing documents:
- 1.
AI-19 (Part VI) Modificati w,;:
Permanent Design Change Control Program
- 2.
AI-9 Control of Temporary I rations and Use of Teinporary Alterations Order.
L. PROTECT plant eq-Lpment, AND DO NOT LEAVE equ ment open and vtattended.
M. USE the correct 1 properly, A,
RTURff the tool to the correct o.lon when a task is complete, AND T.'
DO 11o AUSE tools.
04 fON11-121 $Si~~~~~~~~
0x Ir STANDARD PRACTICE CONDUCT OF MAINTENANCE SSP-6.l Rev 0 Page 18 of 65
.1.7 Specific roup Manager Duties and Responsibilities (Continued)
Instrument Maintenance Group Manager C. RESPONSIBLE to the Maintenance Manager for the management of the following areas:
- 1. Maintenance of plant process instrumentation.
- 2. Maintenance of plant security equipment.
- 3. Maintenance of plant nstalled radiation/contamination monitor.
- 4. Instrument Calibration Program (SQE8 Control of Installed Permanent Process instrumentation).
I Maintenance K >D.
- 5. Housekeeping for instrumentation shops and offices.
- 6. Instr. :entation repair and test capabilities.
Support Group.
>nnger IESPONSIBa' to tire aintenance Manager for the management of the following areas:
- 1.
Management of su, rt craft activities (laborers, carpenters, sheet nb tat workers, painters, asbestos, and composite crew).
- 2. Plant protective coain&.
a,.i preservation efforts.
- 3. Thermal insulation ant' fir.. iarrier integrity.
- 4. Scaffold Program.
- 5. Plant Housekeeping and foreign Material Exclusion Program.
- 6. Tool Room operations, Tocl Control Program, and tool maintenance.
- 7. Measuring and Test Equipm.at (M&TE) Program.
- 8. Asbestos removal and control.
- 9.
Executing warehouse prev maintenance.
hi It545 I
I>
4000 (NV'-IZ et'i
- J.. Holland, OPS 4D-SQN SEQUOYAH NUCLEAR PLANT -
INCIDENT INVESTIGATION (II) ACTION ITEM EXTENSIONLOSURE
Reference:
II Item No.
I 9rozO II Action Item Sequence No.
Os-D I request that the referenced TROI action be Lf closed as complete.
I extended to Basis for extensi n god Atm Zy-+e e.4!
S ai te5 IS
)A Ss ?' %o50 Red-
-1 C-a4l&
bs)Q,^;XS So20ft/
I ts penSilt
- > S S
__
I._aa In X ts A r-t0 At C.AA p
J For closure list supporting close X cumentation as required per SSP.12.9 (Attach Copy).
RK.<
4rtii.'i
/,,'5/'9 3 tIGNATURE (II/AC-fION SUPERVISOR) /
'6TE L-1 This extension does not impact nul:ear safe; I
/ Alternate Corrective Action.
- plant operability.
l0 EVENT MANAGER
/
/ DATE PLANT MANAGER OR DEPT MGR FOR CATEGORY 3 T
/ DTE The responsible organization's and Plant Ms. ger's approval (or Dept. Ngr's for Category 3) is required for all action data x'nsions and for closure when action taken if
'ference from the approved corrt ive at t on in the Final Event Report.
Fl GO AZ 4.,
- PMB PL090205/1317
,nuuI ur ItU1YNiCAL 5UPPORT Rev. 2 PRACTICE Page 5 of 35
.3 Technical Support System Engineers Responsibilities Me 1
Qualifications, training requirements, and experience necessary to perform engineering functions shall be of primary concern when assigning system engineering responsibilities to an individual.
GTE 2 In general, an individual assigned to perform as a system engineer will normally be assigned direct responsibility for one to four systems. However, this will be determined by a Technical Support Section Supervisor based on the knowledge,.
training nd experience of the individual and the complexity of the system to which the ndividual Is assigned.
A. Improv. the reliability and performance of assigned system(s).
B. Perform w -owns n accordance with memorandum, SQN-alkdown Program - T nical Support", (S57 900117 800). This ncludes, but is not 1i.
id to:
- 1. Maintaining.
gu'-r presence in the field.
- 2. Initiating orrect action, if necessary.
- 3. Monitoring performance nd. ators to ensure reliable system operation.
C.
Discuss the operation and maintenm e ~,. assigned system(s) with Operations personnel and maiintenance lanners, to help identify any recurring abnormal operational oT
-intenance conditions.
This ncludes review of appropriate I.. id nstrumentation data.
D.
Trend important system paramneters si;6h as flow rates, pressure, electrical voltages, etc., to ident..
deteriorating system performance.
C.21 E. Initiate corrective action,. pior o ; failure or forced outage; such as, providing nput to the Pt -gram for new or revised requirements.
F. Evaluate significant Ohanges or developing trends and recomrend appropriate corrective actions to the 'mmediate Technical Support Section Supervisor.
EC.21 K>
Provide technical assistance t oti r
- ctions in writing/revising instructions, propr'ied Technical Specifications changes, and FSAR revis'ins.
FI 0001.14 WSi/ds
uwrwucl t TECHNICAL SUPPORT Rev. 2 PRACTICE Page 6 of 35 2.3 Technical Support System Engineers Responsibilities (Continued)
H.
Determine potential problem equipment to be trended in the assigned system(s) and provide documented rationale for the determination of the equipment and applicable performance indicators to be trended.
Trending of selected equipment may include device drift found at calibration intervals, conditions found during SI. O, and/or PM performance, status/condition of equipment during system walkdowns, Nuclear Plant Reliability System (NPRDS) and Equipment Management System (EMS) failure
- reports, tc.
Methods available to the SE for use in determini g equipment to be trended includes engineering
- judgemen, Jaz equipment history, and Reliability Centered Maintenanc
'Ri.') methodology contained in SSP-6.51. The information trimed should be used to predict failures resulting in
-reased unit availability, reliability, and performance.
cial trending requirements are outlined in Appendix C. E ad C.3]
NOTE Component Failure Analy:
'sp.-ts (FARs) will be analyzed by a representative from S s.
C.
ea] with component failures across system boundaries.
[C.
-a C.3]
I.
Evaluate NPRDS and F"¢ xre
-.o..s (SSP-6.4) for the SE K....)
assigned system(s) to c<er'Ine. '1) a failure of the device actually occurred or if some hing ee sulted in the condition of the device, and (2) an unaisirable rc;S has developed with the component. The SE review and respc..^e the NPRDS and EMS failure reports shall be completed an d !merited within 30 days of receipt from the Special Pt ects/fre;d; Supervisor.
Undesirable trends will be resorted tc te im; late Technical Support Supervisor.
C.2 nd C.3]
J.
Maintain System Notebooks r accordance with Appendix E.
K.
Perform a technical review uif new or rev ed nstructions, as assigned, prior to implemel ation.
L.
Revise as necessary, any nstruction itt Technical Support is responsible for and is arilic.eble tc
- ir assigned system(s).
M.
Review workplans for ipact on sys
- i. eractions.
maintainability, sys ?n config ra:
.h -ges, and adequacy of testing performed to slidate Ie yst..i performance after modification.
Review (periodically) ork reques.
CRs) to maintain a familiarity of maintenrxce activit; and he p identify and evaluate repetitive f ures of eqLtment.
5145/ids
_.11 wt LrUt W
UWA1 Ci~cHHlICrAL SUPPORT Rev. 2 PRACTICE Page 7 of 35 2.3 Technical Support System Engineers Responsibilities (Continued)
K.)
- 0.
Assist in the specification of post maintenance testing as requested by other site sections.
P.
Review (periodically) the performance of post maintenance testing activities to ensure the adequacy of testing.
Q.
Review Design Change Requests (CRs) applicable to assigned system(s) for need and priority prior to managements approval.
R.
Recommend changes in system design as a result of items identified through the Nuclear Experience Review (NER) process.
This may include contacting other sites or utilities to evaluate items potentially applicable to Sequoyah. Maintain a copy of appropriate ER correspondence in the system notebook.
(Reference SSP-4.4)
S.
Perform investigations on items which are not resolved by routine maintenance activities.
This may include developing
'oot cause analyses or writing and performing special tests to fain adequate information to evaluate system performance at,./or resolve the root cause of a system malfunction.
T.
Assist
- and/or perform the investigation of reportable occurrent or significant operating events.
U.
Develop prograi-,, analyses, and reports to respond to new or revised regulate. w requirements and requests. (Includes NRC bulletins, Generic 'etters, and notices that require technical responses to met re latory deadlines.)
V.
Provide input to annual b2..R updates.
H.
Provide input to plant schedules for system outages, testing and investigations.
X.
Perform roject management functions for outage related activiti1s as assigned.
- 1.
For major maintenance activities as assigned.
- 2.
For section equipment as assigned.
Y.
Evaluate adequacy of system technical information.
- 1. TE cat manuals.
- 2.
Dr gs.
- 3.
St rocedures as assigned.
Z.
Supportidssist administratively.
F OO01 4;
- 1.
INPO/NUREf/Generic Letter review, NER program, etc.
- 2.
Licensee Event Report preparation.
- 3.
Follow up on plant safety committee e z
AC) I.J.
-.....-. _.~-
qouisun1 Ut-U.1
)L CAL SUPPORT Rev. 2 PRACTICE Page 8 of 35 2.3 Technical Support System Engineers Responsibilities (Continued)
AA.
Investigate drawing deviations between design and as-built configuration.
(Reference SSP-2.1l)
BB.
Establish system!component service/life limitations.
- 1. Review operational requirements to dentify most limiting components (i.e., elastomers, solenoids, motors. etc.)
- 2.
Maintain knowledge of system/component life status.
CC.
Perform periodic evaluations of:
- 1. Products of the Corrective Action Program
- 2.
Safety evaluations
- 3.
12st deficiencies
- 4. Drwing deficiencies S. M&I out of calibration
- 6.
TAC(
DD.
Coordlnate/oL;vai.
endor support as required.
EE. Assist maii,,enc.
Ii
)s with equipment testing and craft support as assigne.
FF.
Submit a monthly letter to ir assigned supervisor n K.)
accordance with the fori.
o Appendlx B.
GG.
Ensure all activities which.
-RL y/indirectly affect reactivity management ar. rev,,
-- Reactor Engineering (Reference SSP-12.17).
2.4 System Engineer Certification 2.4.1 Certification Aecord The Certification Record (A.endlx /) is to document completion of training that demonstrates att nmert of the System Engineer Certification.
Completion of chis -
rtiftcation s not required for performance of the duties and respc bilities outlined in Section 2.3.
These duties will be signed to individuals based on the judgement of the Technical Sup-.u: t Superv'sors.
El OOO1d^£ 0514SIlds
STANDARD PRACTICE CONDUCT OF TECHNICAL SUPPORT SSP-8.50, ATTACHMENT I Rev. 2 Page 1 of 3 TECHNICAL SUPPORT MISSION STATEMENT The mission of Technical Support is to provide technical leadership for Sequoyah Nuclear Plant through optimization of system performance and reliability, quality management of reactivity and assigned engineering programs, proactive identification and resolution of plant ssues, initiation of design modifications and technical assistance to the Operations and Maintenance departments.
1th regard to these areas, the following functions are performed. This Attachment may be updated at any time with the approval of only the Technical Support Manager.
I Technical Support Manager Date
- 1.
Plant Performar.
Establish plan p cance monil reliability anci aff.
c by rot analyzing performance o sth acoustical, and mchcanic4 da~a important to nt reliabil.
analyzed, and ir.t results arE degrading performance.
Serve
_1~ identified problems.
Reactivity Hana-lement Loring activities to optimize plant utinely collecting, trending and is thermal, hydraulic, electrical.
for equipment, systems, and components Efficiency.
Performance data is
,o to proactively predict and correct he '-chnical lead in resolving Technical Support, and n particular the Sta-'on Reactor Engineer, has the overall res.ponK'bility for ensu ng that operation of the reactor core s compatible w.'h the cycle cils, and t..
'ie fuel r....iins within its design basis. Rea.Ivit
.-- o.nt omprisL aspects of how Sequoyah s operated, a. "r phikf..y o.eact5..
anagement demands that (Reference SSP-12.17.
A. All planned reactivity hanges ar. conducted n a controlled manner.
S. The effects o reactivity cha> es a.pe nown and monitored.
C. Any anomalous ndication is met witl.
onservative action.
Program Kanagement Directs the development of assigned progra s and provides technical direction for ongoing implementation of th se programs.
Section XI Repair and Replacement
~
Section XI System Test Ctydro's)
Sectlon XI Pumps and Va' es Appendix J (LLRT AND CII T)
Heat Exchangers Fire Protection se Time Test LA.., lance Instrument (TI-54)
Software Change Control F I,,
Unit Performance TACF's
- 1 tves (SOER 6-003) a
.nee
-I
J. H. Holland, OPS 4D-SQN SEQUOYAH NUCLEAR PLANT -
INCIDENT INVESTIGATION ()
ACTION ITEM EXTENS ION/CLOSURE II Item No.
S Z08 0
Reference:
II
-tion Item Sequence No.
I request that the ercr-ed TROI action be L^ / closed as complete.
1--1 extended to Basis for extension/closu:'.
SQ-k For closure list supporting
. )sure doc. mentat-in as required per SSP. 12.9 (Attach Copy).
SIGNATuRE (I/ACTION SUPERV.SOR)
/
AT
-d VP
,fN#f,,,q3
(.I-IS Pk Ut This extension does not* ir.,act n lear s. fety or plant operability.
/
/ Alternate Corrective Action.
PLANT ANAGER I
D TE The plant managers appre.'al is required or all action date extensions and for clc-e when action is different from the a ',roved corrective action in the Fi e__vent Report.
.:VGT FI GOO3! F.
?L025N08--ID 51
L> 2080 TROI SEQUENCE 6:
£IS92080 TROI Sequence # 6 may be closed due to the existing administrative controls.
STD-9.4 Configuration Management Control indicates site Engineering Managers have management responsibility for establishing and maintaining design basis documents for their sites (except for nuclear fuel components; see Nuclear Power Standard 9.1) and are accountable for the quality and completeness of the design basis within their scope."
This requirement has been satisfied with the Radiation Monitoring System Design Criteria (SQN-DC-V-9.0) and its references.
The attached sheets from the said documents support this position.
Calvin W. Burrel Jr./Date Paperer 13195 Vince A. Bianco/Date Reviewer FI 0001 4G
a..
.I STANDARD CONFIGURATION MAHAGEMENTICONTROL Rev. 0 Page of 8
~_ _-_-
This standard establishes the requirements and responsibilities for the documentation of plant design basis and configuration, and for configuration control.
2.0 S
i The configuration management program defined in this standard identifies documented design requirements, ensures the design is properly implemented, documents the actual plant configuration, and controls configuration throughout the life of the plant.
Changes to plant configuration and the updating of configuration documents to incorporate those changes are covered in Nuclear Power Standards 9.2, 9.3, and 12.4.
This st. dard applies to IVA employees and contractors involved in activit.
L..
ffect nuclear plant configuration.
3.0 uSTRumoNs Configuration L lgement is an integrated process that ensures (a) that plant structures.
-s~ems, components, and computer software conform to approved design r re-nts, and (b) that a plant's physical and functional characte.
it ire accurately reflected in plant documents K) and data systems.
3.1 Design Bafli DoCU=en t q
The design basis for ach uclear plant shall be established and documented.
3.1.1 The Chief Eagiiieer, Corporate
..Lteering, provides (a) requirements or criter.
for identifying and scoping design basis ocuments, and
) defines the engineering processes followed in produce design basis documents.
These requirments, criteria, and processes shall:
A.
Ensure that environmental qualification of safety-related equipment is ir.cluded, as appropriate, in the design basis.
B. Establish a define *lesign responsibilities and i
interface controls t fac;.litate preparation, review, approval, release distribution, and revision of documents Irvolvin d n interfaces.
C.
Implement tL appli' able requirements of ANSI N45.2.11-1974 and Rula:ory Guide 1.64, Rev 2, 1976.
A4)
ONli!&M 0939i
WV
~~~~~~~~~~~~~~1~~~~~~~~~~~~^
P Is I STANDARD CONFIGURATION MANAGEMENT/CONTROL Rev. 0 Page of 8 KU) 3.1.2 Site Engineering Managers have management responsibility for!
establishing and maintaining design basis documents for their sites (except for nuclear fuel components; see Nuclear Power Standard 9.1) and are accountable for the quality and completeness of the design basis within their scope. Site engineering managers ensure that (a) personnel who prepare, review, approve, issue, and revise those design basis documents are trained in the applicable established engineering processes, and (b) requirements, criteria, and engineering processes for documenting design basis are implemented.
3.1.3 The Nuclear Fuel Manager manages establishment of fuel coponent-related design basis documents for each site, and is accountable for the quality and completeness of the es!-n basis for nuclear fuel components. The Nuclear Fuel iger ensures that (a) personnel who prepare, review, L.
ve, issue, and.revise those design basis documents are tra, d in the applicable established engineering processes, and (
Requirements, criteria, and engineering processes f.: doc It -g design basis are implemented.
3.2 Plant Configaurat*
rawsItiou The plant configuratio ha.
je established and documented for each nuclear site.
Pl&.
nfiguration includes the physical arrangements and finctionaL ttributes of structures, systems, and components, and th! computer voftv---, procedures, and other documents that effect functionL
-oitrol of those structures, systems, and compcaents.
Each Site Vice Prdsident may, pth t.
Chief Engineer's concurrence, establish an E-. sion L:.
that identifies site items not subject to onfiguratic. management.
-e Exclusion List may include only structures, sys.:ems, or components that are not quality-related and are not d scribed in the plant Safety Analysis Report. When ri xclusion List Is established, the Site Vice President must implement con' rols to ensure that the list is maintained current.
3.2.1 The ite Vce Pt flent shall establish and ensure imrLewentation of ; processes and controls required to accompli&h document tion of plant configuration. These pr3cesses and c t s shall detail responsibilities of and interfaces betwI er involved organizations, and shall provide r:ntrols for tr.Ln mittal of configuration information tween organize ions.
The interface controls shall provide
.or verifying i
quality of information transmitted between K....>
site organizat r s.
4S ~ ~ ~ ~
~
~
~
~
~
~
~
~
~
~
F
~0S nonQ
RADIATION MONITORING SSTEM SQN-DC-V-9._
1.O SCOPE tJScope This document will ultimately include the design criteria for all monitor channels of the plant radiation monitoring system (System 90).
The second revision of Design Criteria SQN-DC-V-9.0 was written specifically to satisfy one of the system inputs to the Restart Design Basis Document (RDBD), revision.
Therefore, it addressed the design basis of those monitors of the radiation monitoring system that are within the restart system boundaries defined in the QN-OSG7-048, revision 5, calculation (reference 6.1.1).
The second revision also included the design criteria of other monitors that were included in an earlier revision of SQN-OSG7-048.
The third vision of Design Criteria SN-DC-V-9.0 includes all criteria of the sec rtvision and incorporates DIM-SQN-DC-V-9.0-3. The third revision expa.
those criteria to include the design requirements for all monitor cr-
%ia needed to satisfy commitments made to the NRC in CCTS No. NCO86-0.
-00:
The criteria for several additional monitors were available and.
o added in the third revision.
A future revision of the ie:
riteria will add the design requirements for all other monitors.: te
- tion monitoring system. Design t
requirements for some of t
- e tors are provided in the first
~'-revision of this design critEria.
To preserve the documentation o these requirements until they are incc--r ed in the future revision, the first revision is enclosed as Atta. nent A.
All regulatory requirements X d TVA cot itments for implementing Regulatory Guide 1.97 Revisio.n 2 have b -.- incorporated by Revision 4.
If a discrepancy is found :c exist bet yee;
-.is design criteria and another Sequoyah Nuclear Plant (SQN) d s teria, the appropriate Nuclear Engineering (NE) department manu J:r(s) shall be notified by memorandum.
If a discrapancy is found exist between this design criteria and any other document where t other document is not a SQN design criteria, then -- iis design criter! shall govern.
.2 System and Component Identification The overall radiation monitoring system i shown in TVA control diagrams 47W610-90-1 through -6, re rence 6.1.12.
I/
F I 00015" 1
PL02SNO4--1932
-z 9
0 goa 6 COMMITMENT COMPLETION-FORM Part I.
CONTROL NUMBER NC0920161001 o77 Originating Document LER 327/92019 Reference S10921207800 RAF to RC Commitment Due Date 12/31/92 (C)
Commitment Statement: TI-18 and -SI-CEM-030-410.2 will be revised to add -ess the ethod of accounting for gas detector chamber vacuum by Dec. er 31, 1992.
E*..
Part I.
X
- tr-tt Completion Information (Use a separate sheet if ados anal space is needed).
A.
Action Taken.
-nM' te Commitment:
B. Reference Docun.entation:
C. Commitment Completion Da D. Commitment and Documentation Couplet actions as noted in documenta icn are commitment.
and Approved. Completion of all Ificient to fully implement the Signature Impl..:aentir.
rgani-u ion:
In-line Independent V cificat: n:
Signature Lead Coordinator:
Date Date Date Forward to Site Licensing upon cc
)Letion.
Part 11I. Site Licensing:
CCTS Updated to show receipt:
Signature IF I ( ^O.57
Y B olland, OPS 4D-SQN QUOYAH NUCLEAR PLANT -
NCIDENT INVESTIGATION (II) ACTION ITf ETENSION/CLOSURE
!ference:
II Item No.
=
l L 0 ?V II Action Item Sequence No.
request tat the referenced ROI action be 2 closed as complete.
L./ extended to tsis for extensoin/closure:
.\\Ge\\
6;o C
t t
a--'Z s
1pt PS -I I[
LV O.
cloPSl-4 r closure liss supporting clos.- e documetation as required per SSP.12.9 (Attacn Copy).
7,4-
IL
- t. %
3 01 MATURE (ACTION SUPERVISOR) /
DATE
_/
This extension does not impact nulcear safety or plant operability.
_ A ternate ft
&AER Corrective Action.
Z R;
- Wet, Z,
l s
W-PLANT MU&RAGER OR DEPT MGR FOR CATEGORY
/ DATE responsible organization's 4fld Plant Manag:r's approval (or Dept. Mgr's for gorv 3) is required for all action data ex en3 0ons and for closure when action if nce from the approved corrective ac'ic in the Final Event Report.
taken F 000150 3205/1317
'I 't; SLt~llt p=%
LJj.
OuLpullms LUIiw1X)usIL r1lI Monitors RM-90-106/112 and RN-90-130/131 SQN 1L2 PREPARiNG ORGANIZATION KEY NOUNS (Consult RIMS DESCRIPTORS LIST)
MINE Dept.
Rad Monitor, STP,FENCDOSE,COROD,Setpoint BRANCHIPROJECT IDENTIFIERS fac htam She cak"uar t issud. p rwt fttw ongka(RO1 RIMS -rauon nunbew k filled i.
i9 I rRe Io RIMS' ue)
RIMS actesson number SQNAPS3-116 RO APPLICABLE DESIGN DOCUMENTIS)
I N/A R_
SARt SECTIONISI UNID SYSTEM(S) 11.4 NtA R _
Revision 0 RI R2 R3 Safety-elated?
Yes 0 No J ECN No. low endte Not A.patb))
N/A N/A Statement of Problem 4%n:_`_
/
Determine the adequacy of 04CK~s existing Tech Spec setpoints ee. C.
for RM-90-106/112 and Rev iee RM-90-130/131 for limiting K
6 t 145-f doses to 10CFR100 and GDC 19 Apepqed
," i(-
during a small LOCA with D_
containment purge Ust all pages. 4 y this revision.
- Lst all esdeletce 3~by this revision.
z 8 _ List a pes changed by this revision.
Abmract These calculations contain an unverifieo tssumpioni CautKion O s* cw ent%
that must be verified later.
Yes 0 No Qi aflu Yes No ui-~
This calculation determines the adequacy of the existing SQN Tech Spec setpoints for containment (M) monitors RM-90-106/112 and RM-90-1301131.
his was done as part of the corrective action to the radiation monitoring incident investigation reported-in 11-S-92-80. The subject monitors initiate a CVI upon detection of high radiation in CTM atmosphere and purge exhausts. This safety objective is specifically applicable to a small break LOCA event concurrent with CTK purging.
fl OOO15j Source terms for the small LOCA consisted of a release of 00% of the RCS activity with a factor of 10 iodine spike. Total release to the environment consisted of CTM leakage (0.25%/day, first 24 hr, 0.125X.day, thereafter) and CIM purge (14000 cfm of lower CTM, case 1 and 14000 cfm of upper CIM, case 2).
' was used to calculate radioactive purge concentrations at various times K
t-small LOCA. These concentrations were compared with those in Table 3.3.6 of the SQN TSs. FENCDOSE and COROD were used to calculate the control room and offsite doses due to this event to determine compliance with 10 CFR100 and GDC 19.
Results of this calculation are given below.
Tota pages RO: 3G 0
MAcralim and dtore catcuLutsons *n RIMS Sne. Ctev
TITLE BASIS FOR DETERMINING AN ACCEPTABLE SET POINT FOR THEI PLANT/UNIT SPENT FUEL POOL RADIATION MONITOR SET POINT Sequovah 1 & 2 PREPARING ORGANIZATION KEY NOUNS Consult RIMS DESCRIPTORS LIST)
Nf/M/NFSQFP Spent Fuel Pool, Monitor Set Point P- 'INCH/PROJECT IDENTIFIERS Each time these calculations are ssutd. ptrers murt ensure that thi original (ROI RIMS eccevton number Is filled in.
J TI-RPS-181 Rev (for RIMS'use)
RIMS accession number RPS-3147 RO A 4-207qS0 2l,,
No 41206 237 APPLICABLE DESIGN DOCUMENT(S)
S 1 900212EOO 7
9 0 0 1 1 6 0 07 2QN-DC-V-9.O SAR SECTION(SI UNID SYSTEM(S)
N/A
_90 R_
Revision 0 RI R2 R3 Safetycelated?
Yes (2 No 0
- CN No. or indicate Not Aoplicable)
N/A N/A Statement of Problem
)repared M. K. Brando Determine the relation-4c ship between the dose W. M. Bennett rate seen at the spent Reviewed fuel pool radiation F. A. Koontz, Jr z_ r_
_r _
monitors (RE-90-102 and
%pproved 103) due to a G. E. German
____postulated fuel
)ate handling accident and
_12/6/84
//
- a.
the resulting unfiltered all pagesadded
- 10.
offsite thyroid e.
hs revision.l List alt pages deleted N/A by this revision.
al List all pages changed by this revision.
,W ra
/
77
.a U PLA These calculations contain an unverified assumption s) that must be verified later.
Yes 03 No QV An analysis was performed to determine the re the spent fuel pool radiation monitors ME9G-handling accident and the resulting unfilte". v his analysis is to determine if a h gher set po.i.
justified.
nship between the dose rate seen at rnd 103) due to a postulated fuel te thyroid dose. The purpose of/
.r the above monitors can be
/o Occasional isolations of the auxiliary buil.d g ventilation systems have occurred due
- o "spurious" monitor readings in excess of t e current set point,..WmR/hr.
It may/';
)e possible to avoid these spurious' isolations by increasing the monitors' set
'oint without raising the offsite dose to an unacceptable level from releases which o not trigger isolation. Failure to isolate the auxiliary building will affect only he thyroid dose since the auxiliary building gas treatment system (ABGTS) filters nly ffect the odines.
From the relationship between the monitor reading and the nf 1 ed offsite thyroid dose, the accepta ility of a set point can be assessed on
-Pki.is of the offsite radiological conse u nces of a fuel handling accident.
This ysis was requested by NUC PR.
FI (100 5;f.
VUR1INATUN 0E MAIN CONTROL RcN INTAKE ONITOR (047-90-12S.126) SETPOINT IPlant/Unit I SON I 2
eparing Organization IKEY NUNS (Consult RIMS Descriptors List)
NE/NEB/APS3 I RADIATION ONITOR, MHA LOCA, FHA, CONTROL ROOM anch/Project Identifiers Each tifre these calculations are issued, preparers mut ensure that the loriginal (RO) RIMS accession nrbe ir-jilled in.
'9S3-53 I Rev (for RIMS' use)
RTfS ACCESS10N NlJMBFi IRO 870727F0Q13' 145. 87 0.5-3 0 2 3 licable Design Docwient(s)
I i
L&R______________
I I R I UR Section(s)
I UNtO Systen~s)
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Yes (xl No ( )
.R N No. (or Indicate ot Applicable)
IStatarent of Problem Ce Xp
^Zzie;
!I Determine if the 400 cpn setpoint for the main control room air intake onitor (0-RI-90-125,126) is acceptable.
J A
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E FOI4 jList all pages added TA I'-tjby this revision IList all pages deleted 1by this revision jIRED Itist all pages changed lbv this revision I
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I unver. ed. ssu I~
- TRACT These calculations contain an Onts,)Ithat mtst be verified later.
Yes ( )
No x)
, documentation had been found to support the setpoin,' V.
'n in the S ech Spec 5.3.3.1 for the main ntrol roon air intake radiation wmnitors 0-M-90-I25,!26)
'etectors were installed to protect the arators by isolating the control room in the event an acicant
' significant aounts of radioactivity.
is calculation is to determine if the current setpoint of 400 co.....
- .jinute (cpu) is acceptable.
procedure for this calculation consisted cf several t..
The first part of this calculation determined he nt rate expected at the beginning of a LOCA to see if
. is value was greater than the setpoint.
The radio-ive releases due o a axinmw hypothetice. loss of cool i-accident (l-LOCA) were taken fron the three ti ie
?rvals fron a nodification of a STP run found in SNArS3 152.
These intervals were chosen to be representative Ihe accident spectrm at the beginning of the LOCA.
second part of the calculation dterr'
- the count r te the beginning of the fuel handling accident (FHA).
was done in the same menner as the first part.
Th. activi ',s were taken from the first three tino steps of STP run from GEWAL-O08.
I ID 0-1.5 '.
inu i page 2)
- ofilm and store calculations in RIMS Service Cent
. rofilm and return calculations to: F. Taylor IS, SL 26 C-K L. Jones, (WE DSC-OU 3, SH Microfilm and destroy. ( )
Address:
WID 0222 C-J NhE 1-5152(
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NSTRUMENT
- 1. AREA MONITOR
- a.
Fuel Stora'ge Pool Area
- 2.
PROCESS MONITORS
- a.
Containment Purge Air
- b.
Containment
- i.
Gaseous Activity a)Venti lation Isolatifon b)RCS Leakage Detect ion ii.
Particulate Activity a)Venti lation Isolation b)RCS Leakage Detection
- c.
Control Room Isolation RAD [AT ION MINIMUM CHANNELS OPERABLE 1
1 1
1 1
1 JrTORING INSTRUM APPLICABLE MES 1, 2, 3, 4 6
ALL MODES 1, 2, 3 & 4 ALL MODES 1, 2, 3 & 4 ALL MODES ENTAT ION ALARM/TRIP SETPOINT
< 200 mR/hr
< 8.sxlc-3 iiCi/cc
< 8.5x0-3 PCt/cc N/A
< 1.5x10 5 Ci/cc N/A c 400 cpm**
MEASUREMIENT RANGE_
IO 1 -104 mR/hr 10 - 107 cpm 10 -
107 cpm 10 -
107 cpm 10 -
107 cpm 10 -
107 cpm 10 -
107 cpm ACT I(
26 28 28 27 20 27 29 ME SU EM N
I Vithf7uel in the storage pool or building
- Equivalent to 1.0 x 10- 5pCI/cc 0
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STT'r C-ORRECTIVE ACTION ANAfiER SQN SEQUOYAH NUCLEAR PLANT -
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REFERENCEs II ITEM NO. II-§-92-080 I
No. VII.B.5 II ACTION ITEM SEQUENCE NO.
TROI ITEM 09 I REQUEST THAT THE REFERENCE TROI ACTION BE; CLOSED AS COMPLETE.
EXTENDED TO:
BASIS FOR EXTENSION/CLOSURE:
Setpoint and Scaling Documents (SSD's) have been issued for the Radiation Monitors with lch. Spec. setpoints. The SSD's are as follows: 12-R-90-106B, 1,2-R-90-112B, 1,2 R-90-130, 1,2-R-9D-131, 0-R-90-125, and 0-R-90-126.
LOSURE, LIST SUt: ORTING DOCUMENTATION AS REQUIRED PER SSP 12.9.
ATACH COPY ).
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NATURE (I/AJION SUPEvrsop)
DATE PROJECT ENGINEER
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TENNESSEEVALLEY AUTHORITY D(VISION OF NUCLEAR ENGINEERING SEQUOYAH NUCLEAR PLANT SE~ s J T AND SCALING DOCUMENT o0 NSTRUMENT LOOP NO.
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TVA TENNESSEE VALLEY AUTHORfY TENNESSEE VALLEY AUTHORITY DIVI1ON OF NUCLEAR ENGINEERING SEQUOYAH NUCLEAR PLANT SET rt. 'NJ MD SCALING DOCUMENT FC. INSTRUMENT LOOP NO.
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99
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KLa. Holland, OPS D-SQN SEQUOYAB NUCLEAR PLANT -
INCIDENT INVEST-GATION ()
ACTION ITEM EXTENSION/CLOSURE
Reference:
II Item No.
9 C 8 II Action Item Sequence No. _ illz I request that the referenced TROI action beN c-sed 85 complete.
L
/ e:ended to Basis for extension/closure:
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r cosure ist supporting c sul-e documentat-'on as -equired pe-SSP.12.9 Attach Copy).
SIGNATURE (IIACTION SUPL.VISOR)
/
/
/ This extension doer not impact nulceer safety or plant operability.
L/ / Alternate Correctivv Action.
0 7
, i4 Qa VENT MANAGER
/
DATE I
PTW-'IT eIANlAGER, OR DEPT MGR FOR CATEGORY 3 I DATE
.he responsible organizatian's and Plant anager's approval (or Dept. Mgr's
'ategory 3) is required for all action data extensions and for closure when s difference from the approved corrective act;.cr in e Final Event Repoirt.
for action taken PMB L090205/1317 pjof~ A91 J 4 19 S
F1 OGO
.. niwsw w.,,
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rflIYNInm, - zZ; Izt; Setpoint LN1IUNuT SQN 12 PREPARING ORGANZATION KEY OS (st RIS LST)
DNE/NEB/APS3 Radiation Monhor.MHA LOCA, FHA, Control Room Ij5rAr#AV-MUM; IDENTIFIERS 3QNAPS3-053
-I1 EaCh1Nue these Cak8iOnWe ls prppwe mutens theax onil (R) R)JIMS acoessct number is Med kt Ray (bf RIM'S use)
ROAS accssion number
-,~~~
1870727F0O13 B45 870530 238 nO APPLICABLE DESIGN DOCUMENT(S)
P1, NWA R2 SAR SECTION(S)
UNID SYSTEM(S) 11.4 9
YT0(S Revision 0 RI R2 R3 sety retaloos Yes [j1 No [
ECN Nm (or ucdcate Not Appacale)
N(A WfA pparea l1 U.C. Berg Deterine if the 400 csetpoint forthe Ctwx:ed
(
3 main control room air inake monhor W.M. Bennet
(
(ORM-90-125,126) is accetable, exclusive Reviewed Jinstument and san; rcmades. Fut KD. Keith-Jr.
determine the safety it for the subject Approd monitors.
F.A. Koontz Jr.
4 Date W87
_1___
Lit af pages added See Rev Lo by t vison_
IOUl Ust a pges deed See R U9 B=by s
Ltal pges chongd t
o o hese camziabons contain unved ssuptios(s)
Calculan cnams Spew requ._mrnts o thas be vefed laer.
Yes O No fimong cndcon Yes No No documentation hd been found to support the selpoit value given in the SQN Tech Spec 3.33.1 for the main control room Intake ntors (0-RM125, 126). These detectors were tlled to protec the operatons by isolatin the control room in the event an accident released snifct amounts of raroactivity.
This caluion is performed to deten-ine if the culent sotd o 400 counts per mknute (cpn) is acceptabie, exclusi of instrment and samplmg inaccuaies. R is performed to determine the safety unfd for the suect mnitors.
The procedure for fis caulation consisted of several pats. The first part of s alculation deterrined the count rate expected at e beginin of a LOCA to see if this value was greater than the se~pont. The radio-active releases due to a maximun hypothetical loss of coolant accident (MtA-LOCA) were fm the ii time intervals from a modion of the StNAPS3-067 R STP un. These Intervals were chosen to be representative of tf accident spectrum at the begirning of the LOCA.
I, xd part of the callaton detemined the count rate for the beginning of the fuel handing accident (FHA).
T1h 4 as done in the same manner as te first prt The activities were taken from the first ree te steps of the STP un from GENNAL3-008.
(Continued on page 2) 9LI F O il A 19 "-
0 taoflim and store caladions in RIMS Se Cn
Page i SQNAPS-053 R1
/
I i
K he third part of this calculation determined the control room operator doses for the entire duration of a MHA-LOCA as if the main control room never isolated. The ratio of the 10CFR50 Appendix A GC 19 limit of 30 rem inhalation to this dose became a normalization factor. This normalization factor was then multiplied by the release during the 30-46 sec interval to obtain the normalized activity for which the count rate could be determined. The count rate determined in this way gave the initial average count rate at which the operators would receive 30 rem inhalation for the duration of the accident.
The current TS setpoint value of 400 cpm for the MCR intake monitors is sufficiently adequate to protect MCR operators in accordance with the 10 CFR 50, App.A, GDC-19 criteria.
Further, the following safety limits have been determined for the subject monitors in terr-of MC4 intake air concentration (uCi/cc) and MCR intake monitor coun'- -ate (cpm):
0-RM-90-125, 1..
S...ty Limits:
MCR Intake Air Con.
1.81E-3 uCi/cc MCR Intake Monitor 4.:14E4 cpm KXtal Pages R: 42 Fl OOO19`
i.-vI' 9 Ji Iw
,.)r 1 &
RM-91061112 an R.90-1301131 PREIPARING ORGANIZATION KEY NOUNS (Const CCAS LST)
MNE Dept Rad Monitor, STP, FENCDOSE, COROD. Setpoint BRANCH/PROJECT IDENTIFIERS Each ie mwe calciam ar ssed prepay must ensre VW the *uigr (RD) RIMS acsim nt nbt Is Slbd h W,~~S3 116 Rev (kor RIWS use)
RIMS accession number K)APS<l 16 RO 930419G0001 B87 930416 002 APPLICABLE DESIGN DOCUMENT(S) 1 NIA R2 SAA SECTON(S) IUN SYSTEM(S) 11.4 OVnOf O I
Rit R2 R3 Safety reaited?
Ye6s E Nojn I
4 IuN No. (or i001cal NOt 4!NKM)
WA
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VOepAUU RR M=o CheCkod.f M.C. Berg X -,.
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Sment o Prowem Detemrine the adequacy of exsting Tech Spec setpoints, exclusive of inst ent and sanorK inaccuracies, for RM-9G-106M112 and RM90130/131 for mbVt doses to I00FR1X and GDC 19during a maflLOCA fth containmw purge. Estabish safety Ergs for i the subject monitors.
tveweo M.C. Berg
- ApPrOV, R.H. Bryan
, _4 I
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(V/ I /,
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Date 411693
,..~~~~~~~~~I us. ILi'.t pagesedded bw-V OfiS rvasion A
el peges deleted swvn LSpa a
ges changedc a by Vis nevma-Sea RevLog 4-See Rev Log
-4 See Re Log I
I
_ _~~~~~~~~~~~~~~~~~~~
These ctiahions contaka woif s
tios) that uml be veified laser.
Yes []
No Cakulation ctak spe muiremerts or init conditons.
Yes El No KiI Abs Ths cakculaion det inesfthe adequz 'c.
exsng 61N 1 ach Spec setpoi, jo dainm (CTM mnitors RM-9-106112 ad RM-90-1301131 e).e
.^l r f inrwit a :d sampng kiacrad This wasdone as aP t oective acbon tothe radution mnor indent investi si.1 repoted IIS92-8. RI was pedomed to establsh safety mtsfore sutea ontos. Thee montors i te.A V upon thedetection o igh r o
n CM atmosphere ad p exhausts.
i safset obcive is Spi caya cable to a smalbrk LOCA event xcret N CM pUngt.
3ce terms fo the nall LOCA comsisted of a reles ofIC Y of the RCS acviy wfith a factor of 10 one sike. To ease to the erinmen consied of CT
!ea e (.25A.y, frst 24 hr. 0125%ay, theraer) and CTMp e
14000 cfm of oweCTM, case I d 14000 n of upper C M -
2).
TP was used to cauate rdioachve purge a. wentatio*
t auiou.
t-s post-small LOCk hse concentratIs ere '-aed with those i Tab 3.3.6 of tw
'TS.
t 4 OS£ *nd COROD were used t ca te the con
)on \\
aftite doses due to event to det..
e con anc vih 10 CFR 100 and GDC 19.
As of ftis calcukaion are givn below.
Total pages O: 36 j
Tota pages R: 40
.00'A a
IT60
.- a
) Mcrogm and store calcations in RIMS Ser
SQNAPS3-11 6 Pag 2
ABSTRACT (CONT'D)
? SON TS setpoints (8.5E-3 Ci/cc, NG and
.5E-5 pCl/cc, Particulate) for
\\;
>RM-90-106 and RM-90-130 monitors are sufficiently low to init:te a tely
,-tI during a small LOCA with concurrent Lower CTM (LCT4) purging (Case
).
This event resulted in doses that are 3mall fractions of the 10CFRIO0 and DC 19 dose criteria.
Thus, the TS etnoints have substantial margin to limut offsite and control room doses to ithin the dose criteria of 10CFR100 and GDC 19 for this event.
For a small LOCA d
.ng UCTM purge (Case 2), the SQN TS setpoints for a CVI will not be exce. ed for the noble gas channel of the RM-90-112 and R-90-131 moritors. The max m NG concentration attained during this event is 5.88E-3 3lCi/cc vs.. the 8.5 3 --i/.cc T setpoint, t a factor of 0.69 below the TS setpoint. However,
<'.e R
ses due to this event are small fractions of the 10CFRl0 and GDC 19 _
-.:iteria.
RI Thus, although the SN TS noble the setpoint is sufficiently lo maintaining offsitc and cont.:vl
--iteria of 10CFR100 and GDC 19.
03
-tpoint is not exceeded for this event, o.ovide a substantial margin in
. om cnerator doses to well within the dose calculation also demonstrates that he for the particulate channel of the 'M-9%
'2 with UCTM purge. Therefore, tne curr t p ic low to ensure a timely CVI during th eve:..
In summary, all of the current SN 73 CTM mor._
ensure compliance with OCFR100 and (DC 19 I TS setpoints will be exceeded i..tor during the s.l iOCA
-e setpoints are adequately I R setpoints are adequate to This calculation also establishes t.
fll.
- -..g
- safet, imit3 for the utect monitor5n:
- 'FETY LIMITS s
I Monitor RM Safety Limit (
- )
Safety Limit (cp) 106 Noble Gas 1.20E+0_
4.49E+7 106 Part.
1.2E -.
_l.09E+l0 112 Noble Gas 9.71E-..
3.___ OE+6 112 Part.
Lt.l1E-3 1.07E+9 130/131 Noble Gas
?.,71E-2 I
El o COff v
TITLE BASIS FOR DETERMINING AN ACCEPTABLE SET POINT FOR THE PLNTINIT SET FUEL P RDIAtON MONTOR 5 IITT SET POI_1
& 2 PREPARING ORGANIZATION KEY NOUNS (Consult RIMS DESCRIPTORS LIST)
NEZM4/NFMSOP ISPent uel Pola onitar et oint
'AFV6MJ11JJWJr.%;_J IPWiTIFFER5 EN"b tu" e*1 CItl~
S 1U"v, Wperlf nws%5 esum thUs the originallROI RIMS Ve0on hlumbtt I II11 I.
TI-RPS-181 RPS-3147 tor FilMs' u#)
RIMS m"Won number I
MRQNM~~~~
~u~o ume RO ftA ftetMA~ft armim ok^vw APPUCABLL DESIGN DOCUMENTIS)
RI 900212E 00116 007
_8b ?*'D-V-9.o0 I
SAR SECTIONISJ UNID8YSTEMJ NA90 R_
evision 2
A s
Y s
H 0
ECN o. for ndlate Not AppIckb.)
NZA
!!LA l
Statmmn: of PtobIem Prepared Determine tho relation-M. K. Brandon ship between the dose Chocked Y.
. Bennett rate seen at the pent Reviewed fuel pool radiation F. A. KoontzJi.
Jr ntors (RE-90-102 and AppOMVed 1o3) due to a G.. E. Ge an-postulated fue1 Daw_
handling accident and 12LXL6+/-.84 f
/
the resulting unfiltered Listalpsaddd 10.1, offsite thyroid dse.
by this revision.
lt all age, delieted by this revision.
__a____
by tis rvlitcn.
A~~~~~~~~~~-
J/' 1-The calculations contin an unverflIed wurmpilon~s) that Must be verifid later.
Y" 0 N c An analysis was performed to determine he elationship between the dose rate seen at the spent fuel pool radiation monitors RPE-gO-102 and 103) due to a postulated fuel handling accident and the resulting unf itered offsite thyroid dose. The purpose off1
&his analysis is to determine if a hg, r set point for the above monitors can be justified.
Occasional isolations of the auxiliary bu ding ventilation systems have occurred due to spuriousu monitor readings in exces -* the urrent set point,.W'mR/hr. It may(,
be possible to avoid these spurious" s( ations by increasing the monitors' set point without raising the ofsite dose t an unacceptable level frog releases which do not trigger isolation.
Failure to solate the auxiliary building will affect only the thyroid dose since the auxiliary building gas treatment system ABGTS) filters
^ely affect the odines.
From the rlat onshi p between the monitor reading and the
,filtered offsite thyroid dose, the acceptabiltty o a sot point can be assessed on
'-..ne basis of the offsite radiological corsoquences of fuel handling accident.
This analysis was requested by HU PR.
i 0
MirflenaiociieoninRM IncCee.
ttm n
¶t.0
.,eno--7 wkroffim tftd ITAVOY. 0
TI -RPS-181 RIa
- K'act (continued)
The postulated fuel handling accident evaluated was based on the parameters presented in Regulatory Guide 1.25.
The computer code STP was used to decay the activity in a fuel assembly for 100 hrs post shutdown. The photon spectrum also generated in the STP run was used in a modified version of QAD-P5Z which determined the exposure rate (mR/hr) at the monitors as the activity released in the accident rose to the surface of the spent fuel pool.
The unfiltered offsite thyroid dose due to this release was calculated by hand.
The ratio of the monitor reading to the unfiltered offsite thyroid dose is 20 mR/hr/REM thyroid Based on this ratio a set Lint as high as 50 mR/hr would bersl sufficient to meet ANS 51.1 standards (.e.,
10 percent of 10CFR100 limits of 30 REM thyroid).
The computer output is stored on micro fiche Nos. TVA-F-H-501 and 502.
K>
K>F!
01,e I-N
J. R. olland, OPS 4D-SQN SEQUOYAH NUCLEAR PLANT -
INCIDENT INVESTIGATION (II)
ACTION ITEM EXTENSION/CLOSURE
Reference:
11 Item No.
Z/-Ii% -
y II Action Item Sequence No. _
I request that the referenced TROI action be closed as complete.
LkY extended to
///3o/14.
Basi for extension/closure:
X~lzr v/ /tte /'
/v"<e 5- -P '14 t-C o A.-
/
For closure list spporting closure documentation as required per SSP.12.9 (Attach Cop SIGNATURE (II/AC'TION SUY ISOR), DATE
/
s extension does not impact nuclear safety or
-- os x§oq
,operability.
/
I Alternate Corrective A Add$
RESPONSIML Oft-"'
ion.
I
, i-I DATE KGR FOR CATEGORY 3 The responsible organization's and Plant Managek.4proval (or Dept. gr's for Catego ts required for all action date extensions and for closure when action taken is L ferent from the app ved corrective action in the Final Event Report.
JHH:PMB F 1 O0i'j c
1159y
2 Eolland, OPS D-SQN SEQUOYAH NUCLEAR PLANT -
INCIDENT INVESTIGATION (II)
ACTION ITEM EXTENSION/CLOSURE
Reference:
II tem No.
If5 O)
II I 'tem Sequence No.
I request that the reft nced TROt action be II closed as complete.
//I/
extended to
/2&
2
%% pVr^-\\\\* wI 6 Vs4 Basis for extension/clocure:
!6 e______
k.'
- p n
/2;_
A
-. e-gin211A V-
-talc.,4 -.-
For closure list supporting...
d-cun utation as required per SSP.1Z.9 (Attach Copy).
IGNATIURE (II/ACTIw SPERVISOR)
D/.
Of
'_V This extension does not impac nuclea s-ty or plant operability.
fJr
,4V Alternate Corrective Action.
ESPONSL E
"5.
I DATE PL AJ.
- ORACAAEG OPF'. MGR FOR CAEGORY 3
/ /Z-J /
DATE ie responsible organization's and.
.c g.r's approval (or Dept. Mgr's for Category
)
"quired for all action date extensio.. aid for closure when action taken is X£1 t.C from the approved corrective actiou ii the Final Event Report.
.PBMB 59t
~~~~~~~~~~~~~~~~~~F f)
R. Holland, OPS D-SQN SEQUOYAH NUCLEAR PLANT -
INCIDENT INVESTIGATION (I})
ACTION ITEM EXTENSIONICLOSURE
Reference:
II Item No. -TrS Z OE2 II Action Item Sequence No.
097 I request-that the referenced ROI action be
/
closed as complete.
L5 extended to 7/?
3.
Basis for extension/closure:
For _
c o e
s rnc For closur list supporting closul-I documentation as required per SSP.12.9 (Attach Copy SIGNATURE (IU/ACTI0N UPERVLSOR)
DArE vZ h
4W A
a-*47-6~7w -^
/ This extension does not impact nuc.ear safety or antw erability.
t7
.9
.. 3-
-'%' Alternate Corrective Act:.on.
R
~ S0 L O RSPONSIBLE ORG.
6 j/L Z.
DA-1 PLANT MANAGER f
ATE OR DEPT MGR FOR CATEGORY 3 The responsible organization's and Plant Na ager's approval (or Dept. Mgr's for Categor 4s required for all action date extensior s and for closure when action taken s erent from the approved corrective actioi in the Final Event Report.
- U),tvd JH
- PMB I
Fl IOi. 9 1159y 00
IIS92080 V p work performed as a results of the action in Sequencc No. 08 was not adequate to meet the tent of the action. Open a new action reading wDefine the rad monitor safety limits for the rad monitors in Tech Spec 3.3.3.1 and Table 3.3-6.' This action has been coordinate with VAB and is due 611/93. The action of Sequence No. 09 follows this new action and should be extended to 7/2/93. This extension has been coordinated with EMT.
0] ¢>AlqS 3
EI wyoC
,C;
kJ Holland, OPS 4-SQN SEQUOYAB IMCLEAR PN ES SIom/mosu1 INCIDEKT INVSTIGATZON (I)
ACTION IEM Reference.
S UI Item No.
11 Action tem. sgeuence No.
I request that the referancid.TIOI action be
/
1/2 L_
closed as cof.lete.
extended to Fa-I S.
Basis for extension/closare*
p,
>_A x-o~i4I.
+0 4i&
W zae~
E d3 i
X W3/4d0hicL j2t
)
)
-[01 -
i ri For. coiue i;.
zpiidrtinicc re dacmaion as reiied per FSP.12.9 (Attach-Copy)..-
- ~
v SIG;NATRE XJTtsito)/
h-Z This 6itn-doas naact nuclear safety or plant operability.
Alternate:Correcv#. :trq.
The plant aagrs a
Squired fcr all action date extenssons n
closure when actI n l d'£erent from. the approved correctkxe -action in the P'ina Event R sport.- -.
IZ08D2
-'iZKO
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Holland, OPS 4D-SQN SEQUOYAH NUCLEAR PLANT -
INCIDENT INVESTIGAT:ON ()
ACTION ITEM EXTENS ION /CLOSURE
Reference:
II Item No.
S 9
_2 _ _
_0_
I1 Action Item SequencQ No.
f I request that the referenced TROX action be I
J
)>S-
-r, closed as cvnf extended to (3g) 3~~~~~~'tf Basis for eXtL
.or./closure:
L r
cor closure list supporting closuz.
do (Attach Copy).
-at-on as required per SSP.12.9
\\1k 4,Aa SIGIMTE3E (ACTION SUPERVISOR)
IVP SVA.*
.2/
This extension does not impact it clear safety or plant UI Alternate Corrective Action.
-ell-LANT MANAGER DATE The plant managers approval i
. r red for all action date extensions and for closure when action is different o the approved corrective action in the Final Event Report.
HD T
PL-_,08--1D 51 E I AOO.i
Resolving the subject task will require an extensive review of each TECH SPEC radiation monitor setpoint support documentation; see the attached TECH SPEC Tables 3.3-6 for Units 1 and 2. A preliminary review of the supporting Aocumentation for the Control Room radiation monitors
,tpoint indicates a TECH SPEC Change Request in accordance
_sith SSP-4.1 will be required.
Due to the current work load and the approaching UC6 refueling outage, an extension request until 3uly-iG5 1993 is warranted.
I Delaying this task until the said date will not have an impact on plant hardware, procedures, safety, or operability.
A brief discussion for each TECH SPEC setpoint is provided below to support this position.
FUEL STORAGE POOL AREA MONITOR The setpoint for the subject monitor was changed from 15 to 200 mr/hr by TECH SPEC CHANGE NO 104 (S01 850529904).
This change w.
necessary to eliminate spurious actuations of the Auxiliary uilding Gas Treatment System.
The 200 mr/hr setpoint was etermined to be more than reasonably conservative I would not result in a risk to public health and safety.
--efore, the existing TECH SPEC for these monitors will n.
Dissent a risk to plant personnel or the public.
JNTAINENT PURGE AND IS
-S MONITORS The setpoint of these e.
-n.. monitors has been less than or equal to 8.SE-3 ui/cc b
eL n Xe-133 since Sequoyah Units 1 and 2 were originally i,
ed, see Reference NUREG-0658 and NUREG-0789, respectiv"ly.
The Offsite Dose Calculation Manual (ODCM) Lndica s
total body dose shall be limited to 500 r,'yr which is equivalent to 3.32E+5 uCi/sec to comply with NRC 10CF.120...'6 criteria.
The following equation will convert the
'point value from uCi/cc to uCi/sec so a c.4parison can be made SETPOINT PURGE MAX UNXT UNIT VALUE F3OW.ATE CON !RTER CONVERTER (8.5E-3 uCi/cc) (14, 30ft 3 /min) (min ftsec) (28317CC/ft 3).
This equation indicates the conti ent monitors setpoint value is equivalent 5.62E+4 uc. ac r'-ich is approximately 17% of tCFR20.106 equivalent value.
The above equation and eiivalent values of Xe-133 are consistent with the mnithodologies u 4 In TI-30, Manual
)CM) Compliance-Method A.
r,_
q
J.
Fl OOOzo-
CONTAINMENT PARTICULATE MONITORS The setpoint of these effluent monitors has been less than equal to 1.5E-5 uCi/cc based on Co-58 since Sequoyah Units
-nd 2 were originally licensed.
ODCM indicates the total
'5e to any organ shall be limited to 1500 mr/yr which is ivalent to 9.74E+1 uCi/sec to comply with NRC 10CFR20.106 criteria.
In the same manner, the above equation was used to convert the cobalt dose rate for comparison purposes.
Thus, 1.5E-5 uCi/cc equivalent value was determined to be 9.91E+2 uCi/sec, however, the radiation monitor is presently set at 40% of this value (3.96E+1 uCi/cc).
Clearly the current TECH SPEC value is bounded by the release rate criteria specified by the NRC.
It is also noted these monitors are required for Reactor Coolant Leakage Detection as specified by 10CFR50 Appendix A GDC-30 and are not required for adherence to 10CFR20 or 10CFR100.
CONTROL ROOM ISOLATION MONITORS The setpoint of these effluent monitors has been less than or equal to 400 cpm since Sequoyah Units 1 and 2 were originally licensed. The NRC OCFR50 GDC-19 specifies that an operator in the CR shall not exceed a dose of 5 rem to the whole bot"r for the duration of an accident. Using the methodologies n the ODCM, 5 rem was converted to an equivalent val in cpm for comparison purposes.
Based on a 40 hour4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> work wee. tor 52 weeks, 5 rem was determined to be
')ivalent to 386
-om.
This value does not bound the 400 K91 value provided n
- he TECH SPEC.
However, the current setpoints of the MCR monitors were reduced to 253 cpm (4/-
129 cpm due to sensing pressure inaccuracy) by TI-18 Radiation Monitors Re.vision 24 prior to 11-S-92-80.
Therefore, no immediate ations are required.
CONCLUSION The setpoint for the Fuel Pool Storage Area was documented in TECH SPEC CHANGE NO 104 to li"inete spurious actuations of the Auxiliary Building Gas Treatment System.
The setpoint was changed from 5 to 200 r-'hr.
The Containment Purge, Gaseous, and Particulate Monii s setpoints have remained the same since Squoyah Uniis 1 and 2 were originally licensed. The above evaluation has determined the existing setpoints will not resvlt in a dose that will exceed the NRC's OCFR20 or 10CFR10' criteria.
Further reviews are necessary to provide te documentation to support the setpoint values of the above monitors.
This evaluation has also determined the MCR monitors setpoint value of 400 cpm remained he same since Sequoyah ts 1 and 2 were originally licensed. This value was
!ermined not to be bounded by 386 cpm which is equivalent to the 5 rem operator dose spr-4 ied by 10CFR50 Appendix A 91 P I
- nOO3r, p,
mmn
- j; ~
In I
N M
11892080 TROI SEQUENCE 8:
However, the current setpoints of the MCR monitors were reduced to 253 cpm by TI-18, Radiation Monitors.
Therefore, no immediate actions are required however, a TECH SPEC change is warranted to reduce the existing TECH SPEC value.
Don Amos Nuclear Chemistry Reviewer I -- A-
-.,I-e 1 -
Bs E for-k/
Corp. Environmental Protection Reviewer Je! R K. ewtt Technical Supj Reviewer If)4Slu Calvin W. Burrell Jt./Date Nuclear Engineering Daperer
_4{kt&,L*LM - 11?
%IsX, Vinc A. Bianco/Date Nuclea-Engineering Reviewei-F I( O 0 OZQI
C C
m C
-4 p
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-h
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- a TABLE 3.3-6 MONITORING INSTRUMENTATION INSTRUMENT
- 1. AREA MONITOR
- a. Fuel Storage Pool Area
- 2. PROCESS MONITORS
- a.
P...anlui"t Durge Air
- b. Containment
- i. Gaseous Activitv
..,ventilation t nn b)RCS Leakage Detection ii. Particulate Activity a)Ventilation Isolation b)RCS Leakage Detection
- c. Control Room Isolation RADIATION MINIMUM CHANNELS OPERABLE 1
1 I
1 APPLICABLE MODES I
., 3, 4 & 6 ALL MODES 1, 2, 3 & 4 AV./ TsRIP
. OINT c 200 mR/hr MEASUREMENT RANGE 101 -104 mR/hr ACTION R1161 26
~
64 Rl 161
< 8.5x10 3 pCi/cc 10 - 107 cpm 28 28 27
< 8.5x10 3 pCi/cc N/A
< 1.5x10 5 pCi/cc N/A c 400 cpm**
10 - 107 cpm 10 - 107 cpm R16 R116 l 1
ALL MODES 1
1, 2, 3 & 4 1
ALL MODES10-107 10 - 107 10 - 107 Cpm cpm cpm 28 27 29
-With fuel in the storage pool or building
- Equivalent to 1.0 x 10-5pCi/cc RI 161
C11 C.
TABLE 3.3-6 RADIATION MONITORING INSTRUMENTATION
'C S
rn
.0 C
I 9-4
--4 I1.
-P 9.
~-
MINIMUM CHANNELS OPERABLE INSTRUMENT
- 1. AREA MONITOR
- a. Fuel Storage Pool Area
- 2. PROCESS MONITORS
- a. Containment Purge Air
- b. Containment
- i. Gaseous Activity a)Ventilation Isolation b)RCS Leakage Detection ii. Particulate Activity a)Ventilatia-, lsoation b)RCS Leakage Detection
- c. Control Room Isolation APPLICABLE MODES ALARM/TRIP SETPOINT 1
<200 mR/hr MEASUREMENT RANGE 10 1 104 mR/hr 10 - 107 cpm 10 - 17 cpm 10 - 10 cpm ACTION
- 26 28 R 102 JR52 IR102 1
1, 2, 3, 4 & 6 1
ALL MODES 1
1, 2, 3 & 4 1
ALL MODES 1
1, 2, 3 & 4 1
ALL MODES
<8.5 x 10 -3 pCi/cc
<8.5 x 10 3 pCi/cc N/A
<1.5 x 10 5 pCi/cc N/A
< 400 cpm**
28 27 10 -
10 -
10 -
107 cpm 107 cpm 107 cpm 28 27 29 IR102
>J!i 3,
91
-=
'0o
%0 U'
t~4 0~
R With fuel in the storage pool or building
- Equivalent to 1.0 x 10 5 pCi/cc.
0 C,
a.%
.,T lR102
HER ITEM EVALUATION NER NO.
TROI ID (A)
APPLTCABTLIITY EVALIJATION (1)
TITtZ:
_ sr eo 0
&bej-A/
(2) Kevvord(s):
Serr %
r r-e t4-"
/1A4;-,-
94.D12 I
d70 (A)
- Action (3) Actions:
Other (I)
- Info (NA) - Not Applicable (4) Action Priority: 30 days 90 days zOther (PA) - Previously Addressed
/iAC~..
(5) Initial Screening Reviey (6) Screenins Mt Date:- dibc9 Date:
/9/3
/
I A.
Ca M.
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7e rJ s-C (7) Related Documents:
/X R S 2 -
9
& 7 0
1 (8) Incident nvestiiations/NOVs:
Netvork Entry Required:
YES Repeat Event:
FS Key Isues Code:
I I NO I (
1 NO I 'K yL Y,4x)t, (4 8 (9) Significance Code:
5 Lt.
Root Cause Code:
C 8CDj Cl",
c /J 4>A NER Preparer:
Ext Date NER anager:
ExC Z67..
Date
,/FJ3 (B) NER ITEESPONSE SUVHARY (1)
(2)
(3)
(4)
(5)
(6)
NER item is applicable:
Y N
I SCAR/PER nitiated:
Y I N I
Implementation Date:
Item Needs Further evif: YES A
J }
Action:
P I I M[ ]
T ]
Response Summarv:
I (If yes, explain below.)
I O fI l
(7) Item Closed:
YES I NO 1 1 F E x
t
_ _0_ _ a e
.Ext
.Date -
(8) Responsible Manager:
101 8/28/92
TROI I.D. No.:
Cacegory
Subject:
/Vb~u.L
,£V Document No.:
f-S--
0e Responsible Organization:
NER Subject/Site/Responsible Organization:
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NER Des cri on:
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Function:
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o Manufacturer:
60G_3 Key Issues:
References:
t2Xt s2 7 /ZS 2,
f
£.~C 1.2/03/91
TO
- Tose indicated FROM M. J.
echt, Manager, Nuclear Experience Review, LP 5B-C r
u'~ECT:
TRANSMITTAL OF HER ITEM OR ACTION OR INFOkKXTION -
f-S 0O0 Attached is an NER item being forwarded to each of the indicated organizations for action or information as required.
Please provide your
- response, if action is required for your organization, by If action is taken on an HER item that was sent to you for information, notify Corporate NER to correct the assignment.
ACTION INFORMATION Browns Ferry Nuclear Plant, HER Supervisor
[
t 3
Sequoyah Nuclear Plant, ER Supervisor
[
3
[
]
Watts Bar Nuclear Plant.
E Supervisor t ]
Belefonte Nuclear Plant, J.
E.
- ills, OSE-1, B
[
I J.
Rucl.
Xssuraznce. Licensimg ana Tuels:
2.2B.
Jla~h.-^
8B 6A-C C
C I
K. G.
SL -er.
4-C t
]
M. L. TurL
, STC 1I, Sequoyah
[
3
{ 3 Nuclear Projects:
P. C. Pra'vlock.,
4J-C (Sc: __)
I
[
Operations Servicpa:
G.
L. Piser, LP D-C t 3 t
I C.
L.
EI k )
C.
G. Rudson, P
D-C I
[I R. J.
- itts, LP 6.-C E I{
El R. M. McMillan, LP SA-C
[3
]
B. L.
- Wisszman, R 4-C C
I C
I D. F. Goetcheus, BR 5A-C El 1
G.
J. Pitt..,
BR 5A-C
[
1 J.
A. Tear-.'l.
BR SA-C
(
l K. Zi-ernun.,, CST 7B-C 3 ]r Nuclear Materials:
]
[3 Indevendent Safety ngine,.rinq J. D. Robertson, SB 2B. SON 1
3 1
]
D. W. Norwood, PSB-2,
.N 1
I 1
N
- w..
- kiba, PB
, vwB
(
1 t
3 Other; R. H. Rogers, SP 4A-C t I t ]
3
[ ]
G.
. Yelliott, LP 4-C 1
3 t I J. P. Jackson, artsville District Centez 1
]
1
]
cc:
RIMS, MR 2F-C PLNUCBJG/66 1013/92 F I J.OO~r(
103 i41O5
TVA TENN4ESSEE-VALUEY AUTNORITY TENNESSEE VALLEY AUTHORITY OIVISION F M.CLEA EOIE-EZt TPage En l
SEQUOYAH NUCLEAR PLANT SET POINT. A SCALING DOCUMENT FOR 3INSRUMENT LOOP NO.
/_-r 7
I 1
B REVISIN. ROR I
I~~~~
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R2 R4 DATE 1 PREPARED 71 1
I lEVIEWED J
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I TVA TENNESEE VALX ALM.T-CRi' TEuNNESSE VALL£Y A %UTHORTTY OIVyc-'
CLCE-AR SSD0 - F-S9o- 06 rage 3
I ScQUOY NCLEAR PLANTI
$-T PO TNT A SCALING DOCUMENT
,R btruMEckr LCam NO.
A Ie P
r _E 2-F-50-/o&
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APPROV.
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~~~ ~~~~
low V I-n I
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I Ho.
PROBIEM EVALUA ION REPORT (PER)
PER No Revision
° Page of 3
PART A:
INIl IAOR IA Component ID U1 and U2 PASF 2A Plant(s) Orgs. Affected Chemistry 3A ASHE Sec III (
)
and Description Post Accident System(s) 43 and 31 ASHE Sec XI ( )
Samplinp Systems Units Affected Ul and U2 Non-ASME 4A Building Auxilary 5A Ref. Docunents Elevation 706 NUREG 037 and Req. Guide 1.97. Rev. 2 Room No.
Ut PASF and U2 PASF 6A Requirenmnt Violated Item 1: The combined time allotted for samling and analysis should be three hours or less from the time a decision Is made to take a sample.
Item 2:
Equlpment provided for backup (grab) sampiing shall be capable of providing at least one sample per day for seven days following onset of accident and at least one sple per week until accident condition no longer exists.
- Source of Requirenent Violated ____t HNUREG 0131 SA Description of Condition Item #: Boron analysistmelhod cannpt e erformed n three hours or less even with the most proficlent personnel. Approximatelv 25% of personnel are capable of meeting time restrictions of three hours or less excluding B analysis.
Item 2: Humber of personnel capable of performing sampling of PASS and analyses of PASS samples satisfactorily Is not sufficient to provide for the frequency required.
Restrictions on dose levels allowed prohibit repeat performance or smplinq under accident conditions.
9A Initiator Gregory D. aylor Organization Chemistry Date 4/14/92 Phone No.
X8697 SUPERVISOR IOA Confirmed Adverse Condition:
Yes Potentially Affects operability:
Yes (
)
Potentially Reportable:
Yes (
1 ats SCAR Criteria:
Yes ( L )
Assigned Responsible Organization Supervisor's Approval J
§-F4 6
-WRIHP'3-s59 L N st LUMUMb4 /5 AAUWoo Ho
)
If no, a ustification must be. attached.
No (Viz No ( v*1iJ qvs3(9 No 4
4f yes, SCAR Ho. _
Coordinated WIth e_
_ie___
Date
/~
92 O CG 000003 LIV
PER N L/E PROBLEM EVALUAlION REPORT (PER)
R K..>
~~~~~~~~~
~~~~~~~~~~~Re 0
Continuation Sheet Page 3
of 3
PER CONTINUAIION Identify the information that s being continued on this sheet.
(For example: Description or Condition.)
NOTE:
Entries made on this sheet shall be signed and dated.
6A. Item 3:
The post accident reactor coolant and containent atrmsphere samples should be representative of the reactor coolant in the core area and containment atmosphere following an accident or transient.
BA.
tem 3:
Diluted PASS samples to not meet cparison criteria when cpared with hot sample roan results.
Gregory D. Taylor 4/7/92 CG 000004 A A Rmmal 4
,2 18U Mg4yP 12-90)10-913 ft"%uVV.&JLJ6
I SIGHIFICANT CORRECTIVE ACTIOH REPORT 1-1)
SCAR No. S 9 S C A 9 2 OO O 4 Rev. No. 0 1
Continuation heet
.~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~(
CONIINUED NOTE: Entries ade on this sheet shall be signed and dated.
Step o.
Response to Step 6E (QA Verification) submitted 2/24/93 The CAP for SQSCA9ZO0O4 developed July 1, 1992 was Intended to supersede proposed corrective action plan developed ay 14, 1992. Proposed CAP (5114/931 was Interim (i.e., temporary) plan until study/root cause analysis (RCA) was generated. As a result of the RCA SQN Chemistry formulIated a more comprehensive CAP.
Signature i f Title "-16 C-.l 594A (UP 1-92)14-921 as 01/996/11 CG 000005 AAOOO1IZ
._.GNIFICANT CORRECTIVE ACriON R 1E it'T 11gc 2 or_
SCAR No. Slolrld4 I I Rev. No.
5 -ny Rcsponsibi Orgrniznlion 7J(S) O;cticrillcvlcw acquired
[jaY";
No E3'3r-OSON 0"W1Ns
[P"l.
[3Corpor;,
Just iriaitinn (S )
__r__wrc isrs_
__n llcp ir
__Accc_
__-ls
__ork Scrnp
__NA__'
(50 I n Ic rim A c lins GYCs No Se>
Rr.
Woft 1it13LM5 (51)) Itmit Causc Code
__oo_
Cns_
__nclusin (ft +i~tols ;.reA ; yr< ;..- Th 7 Af (SE)
Extent or Conlition VConside 0
r ofcr unil(s))
iC p
- ,E1 I;
t'Sfl Corrccdive~tin (SCII Compiction l)alc (p-B0 9 LR 1
Corrective Action 1'ktn: Concurrence I Approvals Concturrencc Nam~e (int vr Type) ltiliJtls Pafe Approynls Name (Prn orType)
Dulil ate (511) 1irp.rcr N~_
_(SAl)
O A,~
(51)
Supcrviscar
/
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'(S) ~fC.
(
LA' (50)
Mont Mgr. 5CecA PzA (SK) Aion Org (5i.) Action Org SI'E' 6 - Clistire Name Pdinr r Type)
Initials Date (6;) All Actions
(
- Complete.
LI -Ic %
(61Q Responsible Supv. Mgr.
crif icI ion t
kNc I ANI rifi
,.n JJI A
51JTngs Removed IV /t CG0006. _
(6E)
Acrifintion
_0 (61SE SCA Coordinetor NIor g).D VA 19384 Mlr 121 A A OnnlhI~
Sh-.FIC'AN1 C1MUL( 1-'VE CA11E SCAR No.I jjCjFV ACTION REpO, Step No.
Rev. El 4c 6O I
E besge addt CONTINo. -
NUEL)
NOTE: Entries madc on this slct shall CR.
a
.f
'2F 7
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'/A A O O 11 4
- G 000007 I
SIGNIFICANI CORRECiIYE ACtION REPORT SCAR No.
S C A 2 0 0 0 4 Rev. No.
O ContinuatTo Whet -
5F CONTINUED NOTE:
Entries imde on this sheet shall be signed and dated.
Step No.
AMENDMENT OF CAP FOR SCA920004:
INADEQUA1E ETIIOO/EQUIPMEN1 FOR PASS ANALYSIS:
lo be completed by /15/ 92-GDT/CEg.a re
, /
Signature
-71t~tt gr
/f Title f/6lf C -
e, 1 O-.
action item for analysis by Standard Method of Additions. PASS matrix required by NkC for testing may lnterfere with this analysis.
I8>z.-
1Id by 7!
GWJ/CEH.
v
/
7PI" C ¢,,,paiC Signature Title Af6 c//
7U INADEQUATE RAIHING/PRACIICE AND LOW MANAGEHENI PRIORITY
- 3. Incorporate In A-14/ISQN a PASS Qual Card with clearly Vi d qualification criteria.
To be conpleted by 7/30/92-RDB/CEM.
Signature Title
- 4. Establish required performance of 1/2-T1-CEM-043-066.1 within 30 days of each performance of 1/2-PI-CEM-43-487.
lo be copleted by 8/6/92-GOT/CEM.
/
/
Signature/>B/AfC^7/ 5 Title
- 5. Close action tem for PASS NG to be provided with approprlate documentation of training performed.
State 3 hr. requirement not met due to equipment.
Ihis is being addressed by action Item 1.
It1 4 0
.lr
&.0 h'
!--GOT.CEM.
'7 4
C Pe f
,fZ sature f l tS f
iw, litle_
G-.
INADEQUATE LONG-TERM DOSE ASSESSMENT
- 6. Site DNE to provide projected dose over time in a RIMS eo.
This Is to specify projected dose to be received during each day from day 1-7 for operators A and B.
To be completed by 11/10/92-CWB/ONE.
Signature L l
< trj I
cq Title _
__ _A CG 000008 TVA l95 4A (HP 1-92) 14-92]
AAQOO1I.5 VLO30 1/995/3
SIGN IFICAN1 CORRECTIVE ACTION REPORT SCAR No. S S C A 9 Z OO 0 4 Rev. Ho.
0
~ Conliinuitlon hieT-CONINUEU NOTE: Entries made on this sheet shall be signed and dated.
Step No.
AMIENDHENT OF CAP FOR SCA9Z0004:
IHAOEQUATE VERIFICATION OF PASS SAMPLE:
- 7. Perform a study of dilutor valve perfonmance to establish act I dilution factor.
lo be cpleted by 10/l/92-GDI/CEM.
1/
Signature fiB/d i/
Itle
- 8. Establish 0-TI-CEH-260-020.9 to inplement a QC program-based n g"'dnr.auvreo Dy IPO good practice 91-0l9.
be-ew*
sI
-DJB/CE
/
signatur IK /
-tt.-g Title _
Sigl g n
a t
/
Title-iAgk>C. Signature Title
,ege 79 At A, s
rA4A1461-R Signature _
Title r, / / /
t L
/
- 7.
2 Al Signature Title AA000116 CG 000009 I
IVA 19584A INP 1-92) [4-92]
PLO2020I/99 6/4 1/92
v s
rF
-
I
- 1. I- -
_GNIFICANT CORRECTIVE ACTION REPORT SCAR No. I S I a.'E cIr tcAi +/-kolL 0i Rev4.
[
Identify the infornmation tht is being continued on this sheet (or exa=mpc: Dscriprion of Condition).
£C UFr el?
A C W
NOTE: Entries made on this sheet shallbe signed nd dated.
PROPOSED CORRECTIVE ACTION PLAN
,. Complete PASS training for all techs. Verify ability to sample and analyze under three hours.
- 2. Exercise via simulation tale PASS system on a weekly basis to maintain proficiency.
- 3. Maintain capability from shift to shift of having three qualified people at all times with the first option to hold over from the previous shift or call out to a designated beeper responsibility.
- 4. Short term due to equipment nad uancy, modify the appropriate procedures to facilitate boron analyses ti a on by standard addition.
Long term a more desirable alternative based o inimum detectable limits, reduced setup time is boron analysis by plasma spectroetry (a purchase request has been submitted for a plasma unit).
4AL4',Z>
/
Z AL' eG/wos N
I SOuL
-lI-l It:
i Jr..,d 3 AA0001'17 CG 000010
DCNI iM-O4zq Page z S of SAFETY ASSESSMENT DCN
-06299B B.47 Valve Altgnment c1allges:
Addition of the new plug valv (identified as "RC-V-23") via this modification will requir changes to the procedures which govern the operation of th sampling panel.
The changes include (1) identification o the valve, (2) specification of the valve alignment (a.
closed) during all operations except undiluted off-gas sampling, and (3) addition of the new sample procedure.
Revision to these procedures will take place following the modification and will be made in accordance with approved plant procedures.
B.49 New Radioactive Effluent (Liquid or Gaseous)
Release Pathways:
No effluent release pathways are changed by this modification.
The PASF IIVAC will be in service to provide ventilation for airborne releases made within the PASF while sampling the undiluted coolant off-gas. UFSAR section 15.5.3 discusses the release path for Post-Accident Sampling, and this method is not being changed (i.e., See Safety Evaluation part B.3).
During the sample transport from the PASF to the RCL, a release (if any) will be via ABGTS which will be in service with its normal release pathway out the shield building vent.
Once within the RCL, any release will be via the exhaust loads out the service building vent.
All of these release pathways are monitored vents.
CG 00>01 I AI&000fl 8 OG 000011
IPage aL.
of SAFETY ASSESSMENT 17 DCN M-06299B K)J B.25 Water Spray/Condensation:
The modification adds tubing, a tee fitting and a valve interior to the panel and does not introduce new design parameters (i.e., temperature/pressure).
The new stainless steel components are designed to be equivalent to the existing components (TVA Class G seismic I(L)B) (Ref. A.2.b).
The modification of canging Teflon tubing to Tefzell is more than adequate since calculation, reference 2.A.jj, shows Tefzell to be superior to Teflon in all categories (pressure, temperature and higher radiation dose).
- Also, the panel
- supplier, Sentry Equipment Corporation as recommended Tefzell as a suitable replacement F f or the Teflon (Ref A.2.mm). Therefore, the potential for or consequences of spray/condensation is not changed by this modification and is not a nuclear safety concern since no safety related equipment is located within the panel.
- B.26 System Design Parameters:
The additions to the system as proposed in this modification are designed to the same parameters as the system now in use.
(Refs. A.2.d.1, A.2.b, and A.2.jj)
Thus, design parameters are not a concern toIp nuclear safety.
B.27 Test & Retest Scoping-Document (Post Modification Test): The only post modification test required is to assure that the added tubing, fittings, and valves are leak free at the K) operating pressure.
since all joints are compression type, this modification would be subject to in-service leak testing per the Modification criteria prior to routine service (Ref.
A.2.j).
B.28 Chemistry Changes or Chemical Release Pathways: No chemical release pathways are changed.
The existing sampling point for diluted gas sampling (e.g., the receptor for the 15 cc sample vial) shall be used for the undiluted gas sampling.
Thus the same pathway is being used following the modification, and no new pathway is introduced.
Changes to Chemistry procedures (e.g., TI-66.1) will be necessary after this modification, but the changes will be in method of sampling and analysis, not in plant chemical parameters or requirements.
Thus, nuclear safety is not effected and is not a concern.
B.36 Reactor Coolant Pressure Boundary:
A breech in the bypass line installed by this modification is isolable from the TVA Class A piping (Ref. A.2.1); thus, the line and modification do not qualify as Rcs Pressure Boundary.
B.38 Pipe Vibration:
Per the vibrational analysis of this modification (Ref. A.2.c.), pipe vibration will be of no concern to nuclear safety.
CG 000012 AAOOOIL1
SIGHIFICAN1 CORRECTIVE ACIIOl REPORT SCAR No. S S C A 9 2 0 0 0 4 Rev. No.
0 ContiiutTon heet 5F CONIlNUED NOTE:
Entries made on this sheet shall be signed and dated.
Step No.
In reference to item I of corrective actions for SCA920004
- 1. Dedicate one channel of the ion chromatograph for the analysis of PASS boron. lo be completed by 7/15/92-GOT/CEM.
This item cpleted on 76/9Z.
Signature ('r.iŽ%
.w 4 Y.. \\
-S
.t.
71tle V.
B s
Sc^.&
i.1; I
AA000120 CG 000013 I VA 198jA(& 1-92) [4-92]
PLOZOOI99/
S
,IGNIFICANT CORRECTIVE ACTION Rt. ORT SCAR No.
15; 2.s 1 1°1 1 eRel.
i A
Identify the information that is being continued on this sheet for example: fescriprion of Condition).
NOTE: Entrics made on this sheet shall be signed and dated.
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I AA000121 TVA '
AykVX.Nrrn5
'P.ROCEDUR.E oti-TIIE -JO0 TRAINLNG (OJr)
Page o
21 K>
APPENDIX A (Page 3 t 4)
TASK QUtALIFICATION STANDARDS MANUAL Page 2. of SXA= CuZTlcasoK S)J.A=D COVE2S9
- Ae,
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.~~~~~~~~~
- 1
'I0 f;
TVA 40239 (UP 4/92)
Page of 2 AAOOO1Z2 CG 000015
- ...e*. -.
-~d=
Rev.
C1111P'0 023014 PaE of 9 TASK QUALIFICATION CARD FORK TASK QUALIFICATION CARDS MANUAL VOLUME I
PROGRAt Chemistrv RLA f
% QUAL CARD TITLE Post Accident t
5 SAMpling Svsten PASS) Team 5
5 Description of Change tPage(s) Affected t
_r(__
[
e and ink cha ces attach copv of affected oage(s)l and Revision Level
)PEN and INK CHANGE( )ADD Si'ejjdj(
)FORMAIL REVISION
( )DELETE k
A.."
c Initial issue.
Transmitt to all Rev.
x)INITIAL ISSUANCE ( )CHANGE (p
panager Nuclear Training ites.
)PEN and INK CHANGE( )ADD Site Hahager
)FORHAL REVISION
( )DELETE
)INITIAL ISSUANCE ( )CHANGE Training Manager
)PEN and INK CHANGE( )AOD Site Manager
)FORMAL REVISION
( )DELETE
)INITIAL ISSUANCE ( )CHANGE Training Manager
)PcN and INK CHANGEQ )ADD Site Manager AL REVISION
( )DELETE K.TIAL ISSUANCE
( )CHANGE Training Manager
)PEN and INK CHANGE( )ADO Site Manager
( )FORMAL REVISION
( )DELETE
( )INITIAL ISSUANCE ( )CHANGE Training Manager
( )PEN and INK CHANGE( )AOD Site anager
( )FORMAL REVISION
( )DELETE
( )INITIAL ISSUANCE
( )CHANGE Training Manager
( )PEN and INK CHANGE( )AOO Site Manager
)FORMAL REVISION
( )DELETE
)INITIAL ISSUANCE ( )CHANGE Training Manager
)PEN and INK CHANGE(
)AOD Site Manager
)FORHAL REVISION
( )DELETE
)INITIAL ISSUANCE
( )CHANGE Training Manager
)PEN and INK CHANGE( )ADD Site Manager
)FORMAL REVISION
( )DELETE
)INITIAL ISSUANCE ( )CHANGE Training Manager
)PEN and INK CHANGE( )ADD Site Manager
)FORMAL REVISION
( )DELETE
)INITIAL ISSUANCE
( )CHANGE Training Manager AA000123 CG 000016
A. -
V C1111020.023.014 Page 3 f 9 RLA OJT QUALIFICATION CARD POST ACCIDENT SAMPLING SYSTEM (PASS) TEAM Team Members' Payroll Name Plant/Date Social Security Number 1.2.2 SQN 1.2.2.1 1.2.2.2 1.2.2.3 1.2.2.4 1.2.2.5 H032 11033 H039 H042 H047 (609049)
(609050)
(609056)
(609063)
(609064) 1.2.3 BN 1.2.3.1 1.2.3.2 1.2.3.3 1.2.3.4 1.2.3.5 1.2.3.6 G049 G050 D172 A174 A119 A120 (609049)
(609050)
(609056)
(609059)
(609063)
(609064) 2.0 Performance Standard Demonstrating required safety measures and practices, the team, within the three hour timeframe, shall procure and analyze samples from the PASS for accident conditions per referenced site procedures.
3.0 General References*
3.1 BFIT f.
AA000124 CG 000017
t.
SIGHIEICANI CORRECTIVE ACTION REPORT SCAR No.
S S C A 9 2 0 0 0 4 Rev. No. 0 contiiuitToli heet iF_
StepNo.
CONlINUlD NOTE:
Entries ade on this sheet shall be signed and dated.
Step No.
TI S&
I )
reference to tem 4 of Corrective Action Plan SQSCA920004 Establish required performance of /2-T1-CEM-043-066.1 within 30 days of each perfonance of 1/2-PI-CEM-043-497.
his tem cpleted 7/17/92 by Includin? In cments section of 1/2-PI-CEM-043-487 cover sheets that 1/2-7I-CEMH.043-466.1 Is to be perforzed within 30 days and to ensure that WCG has scheduled performance n WCG database.
Completed 7/17/92. GOT/CEM.
Signature/Date I
- r 7-/7-'V?
7itle t2L "
/
4, s csz.,l. A (I
CG 000018 ffiYI~dgj)09t% 1-92)14-923 A ANO OO1Z
I" TO: MIKE GOODSON, CIEMISTRY LAB SUPERVISOR SQN f
K'jFROM: RICIIARD D. BAYLES CEMISTRYINSTRUCIORSQN DATE: 3/3/93
SUBJECT:
PASS QUALIFICATIONS THE FOLLOWING HAVE COMPLETED PASS TRAINING ON THE DATE INDICATED.
ADCOCK DICKIE L. 4122192 CLINE BOBBY K. 51192 CURVIN DOROTIHY B. 411f6/92 DAVIDSON JACK K. 5m192 DEMOTT DAVID B. 3/3/93 FERRY LINDA S. 2/15/93 FINCIIER MIKE L. 5n192 GOODSON MIKE C. 527/92 IIARDEMAN DELORES D. 3/3/93 tlARRIS IIAROLD L. 4116/92 MANN llAROLD R. 2/15/93 MARSTON STEWART D. (INITIAL 7/3/92) 224/93 NIDA DIEDRE B. 2/21/92 NUNLEY RICHARD D. 2/15/93 PAGE BARBARA A. 4/22192 PAPPU THIJMOTHY P. 4/9/92 PARKER JEFF A. 2/24/93 K>
PIERCE JERRY D. (INIlIAL 713192) 2124/93 SEAY RONALD 313/93 TAYLOR DON W. JR. 1/27/92 TAYLOR GREG D. 4/16/92 THOMPSON DAVID M. 1/31/92 THURMAN RODNEY S. 4/9/92 WADDLE CARROL M. 2/24/93 W L] IOITE Cl ARLES R JR. 1/27/92 WRIGI IT WAYNE A. 5128/92 DADE MASSEY HAS NOT COMLETED INITIALTRAINING.
AA000126 CG 000019
5 cA9;Ot)L e&s TO: MIKE GOODSON, SQN LAB SUPERVISOR FROM: RICHARD D. BAYLES, CHEMISTRY INSTRUCTOR STC-SQN DATE: 6-05-92
SUBJECT:
POST ACCIDENT SAMPLING QUALIFICATIONS THE FOLLOWING PERSONNEL HAVE SUCCESSFULLY COMPLETED POST ACCIDENT SAMPLING (PASS) TRAINING CHM005.502. THIS COURSE IS A RETRAINING COURSE WITH 1 DAY OF CLASS ROOM AND SIMULATOR AND 1 DAY OF IN PLANT TRAINING ON THE SENTRY POST ACCIDENT SAMPLING SYSTEM.
DUE TO PLANT CONDITIONS PORTIONS OF THE IN-PLANT TRAINING HAD TO BE SIMULATED.
- 1.
- CLINE, BOBBY K.
5-7-92 SIMULATED IN-PLANT
- 2.
- FINCHER, MICHAEL 5-7-92 SIMULATED IN-PLANT
- 3.
- DAVIDSON, JACK K.
5-7-92 SIMULATED IN-PLANT
- 4.
BLAND-THEIM, ANGELA 4-30-92 SIMULATED IN-PLANT
- 5. PARKER, JEFERY A. 4-30-92 SIMULATED IN-PLANT
- 6.
- LEWIS, WILLIAM RAY 4-30-92 SIMULATED IN-PLANT
- 7.
- WADDLE, CARROL H. 4-30-92 SIMULATED IN-PLANT
- 8. ADOCK, DICKIE L. 4-22-92 SIMULATED IN-PLANT
- 9.
- WILHOITE, CHARLES R. 4-22-92 SIMULATED IN-PLANT
- 10. PAGE, BARBARA A. 4-22-92 SIMULATED IN-PLANT
- 11.
- CURVIN, DORTIHY B.
4-16-92 SIMULATED IN-PLANT
- 12.
- THOMPSON, DAVID M. 4-16-92 SIMULATED IN-PLANT
- 13. MANN, HAROLD R.
4-16-92 SIMULATED IN-PLANT
- 14. HARRIS, HAROLD L. 4-16-92 SIMULATED IN-PLANT
- 15. THURMAN, RODNEY S. 4-9-92 OBTAINED SAMPLES
- 16. FERRY, LINDA S. 4-9-92 OBTAINED SAMPLES
- 17. NUNLEY, RICHARD D. 4-9-92 OBTAINED SAMPLES
- 18. PAPPU, THIMOTHY P. 4-9-92 OBTAINED SAMPLES
- 19. DEMOTT, DAVID B. 5-27-92 SIMULATED STC
- 20. GOODSON, MICHAEL C. 5-27-92 SIMULATED STC
- 21. HARDEMAN, DELORES D. 5-28-92 SIMULATED IN-PLANT
- 22. NIDA, DIEDRE B. 5-28-92 SIMULATED IN-PLANT
- 23. SEAY, RONALD 5-28-92 SIMULATED IN-PLANT
- 24. TAYLOR, DONALD W. 5-28-92 SIMULATED IN-PLANT
- 25. WRIGHT, WAYNE A. 5-28-92 SIMULATED IN-PLANT
- 26. TAYLOR, GREGORY D. OBTAINED SAMPLES 4-9-92 THE FOLLOWING HAVE NOT RECIEVED INITIAL TRAINING: (SCHEDULED FOR 6-29-92, 6-30-92,& 7-2-92)
- 1. MARSTON, STEWART D.
- 2. PIERCE, JERRY D. (RETRAINING 5-28-92 SIMULATED INPLANT)
AAOO0127 OG 000020
Eatp D.
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. 'hurman nley __________
C. Waddle e
C. WIlhoite pr W. Wright rker Uarston erce Watts Bar ay Watts ar 7 8
-W 6 aSCAl L
Post Acci-ient Sampling -
SCAR EQSCA92OOO4 Belot is 'he ccrreczive action to SCAR SQSCA972OOO4 associated with providing
- .-verage in the event of an accident.
,ainzain capability frem shift to shift of having three qualified people at a!:
imes witn the first option to hold over from the previous shirt or call out to a designated beeper responsibility.
K his coverage includes the CSS on duty.
Danny ayles is in the process of training perscnnel on the PS'SS unit and will provide a list of qualified personnel at ;he end f the training eriod.
In the event an individual dces not pass, their uali iwations as an ?.A will be evaluated. S'hift coverage for a nn-qualified ndividuai ill be accomplished by retaining the Analyst on the previcu_ sni:t no has.he l'*,est., mount of overtime.
K>
lnitiu.tei b:
'Th.emjs.ry (.ntrl Etierviscr:
S).,
CG 000021 AA000128
SIGNIrICANT CORRECTIVE ACTION REPORT SCAR No.
S S C A 9 2 O
4 Rev. No. 0 ConTinuatTo heEt -
5F 443 CONTINUED NOTE: Entries rade on this sheet shall be signed and dated.
Step No.
Chemistry requests closing of CA 4 of SQSCA920004, Rev. 0. Corrective Action 3 will ensure qualification and proficiency Is aintained.
C.
1.
0, Ig',gXi,/426z Signature/Date Title__ /t,
1h6 rl-111)"I'll '.
§,
bf:
(--I -1 7
It/
7 Y. 384A (P 1-92) [4-92]
- 01/996/t2 CG 000022 AAOO0129
S52 930309 187 Harch 10, 1993 R. J. Beecken, roB 2 SrQN Due to Operations' desire to minimize risk to the plant, PASS will operated except in Hode 5 until DCN 00608 nnd 00746 are complete.
will provide the ability to obtain PASS samples without entry into not be These DCHs LCO 3.6.1.1.
'.i C. E.
ent Clemistry & Radcon Hanngter Po 2C-SQN VJB:BT cc:
- RIHS, HR 2F.-C Concurrence:
Op rntionls Hnaeer PL020201/3578/2 AA000130 CG 000023
K>
SIGNIFICANJ CORRECTIVE ACIIOH REPORT SCAR No. SQ S C A 9 ZO O O 4 Rev. No. 0 Continuation heet 5F 6 CON1INUED NOTE:
Entries made on this sheet shall be signed and dated.
Step No.
Based n the attached S/SE for OCH i06299, the current level of RA staffing Is adequate to fulfill NUREG 0737 requirements.
Signature/Date
//w Title I /
LZ
.i.
("r (I
I'"
17 34A NP 1-92) 4-92]
u
/996/13 CG 000024 AAOO0131
SIGNIFICANT CORRECTIVE ACTION REPORT, yOj SCARNo. I SIQ CA I 921 0 0
0 4
1 1 Rev. [E Continuation Sheet 5F 6
CONTINUED NOTE: Entrie made on this shet shal be signed and dated.
.Step No.
SQP 920004 SCA Sequence 15 be may closed based on Revision 4 of the A/SE for DCN M06299.
Prior to Revision 4 Section B.55, PASF MANPOWER of the SA stated 35 RCL analyst and 14 RADCON personnel were required to operate the PASFfor the.
30 day period following a design basis LOCA.
This an power was based on the Gilbert/Commonwelth contractors incorrectly interpretation NE calculation TIRPS-146, POST ACCIDENT.
DESIGN BASIS SAMPLING MISSION DOSE.
This calculation only calculated the design basis mission which is the first or worst case mission following the accident.. The above personnel requirements were based on dose rate calculated i, hour after the LOCA as required by NUREG 0737, Some of.the mission team members received 4.95 rem in the design balis mission, while the NRC dose rate limits povided in DC 19 is 5.0 rem over a 30 day duration following the LOCA, Revision 4 corrected this interpretation error by deletihg the above personnel requirements and indicated the. gamma radiation shine from containment decreased approximately by a factor of ten after 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and decreased substantially more over a 30 day period.
A copy of.FSAR Figure 15,5.3-7 was is also attached to further support this position.
Calvin W. Burrell J./Date.
NE-MN Nuclear Engineer Johne Barker/Date'.
Nuclear Chemistry..
A.
AA000132 CG 000025
Cl Co 0
0 0
(9 0
I u-I SEaUOYAIi NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORT DOSE RATE
..... \\
~~~~~~~~~~~~OF
.CONTAINED SOURCES
_11.,-
Y
.Figure 15.5.3-7 I
10 w
U)0 0u' o
w I-U' w
m DOSE RATE ZONE AT 05 HR (MR/HR) 1 2B x 103 2
I.9x 104 3
3.Ox 104 4
1.7x 105 5
2.6xiOi 6
3.7x IO 7
I.IxIO" B
2.7 xlO8 9
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)
ins l)lC.7 PLIt Nu. &QlNzoooq PROBLEM EVALUATION REPORT (PER)
IlcvL-;ion CONTINUATIONSHEEl' I'age 4 or PER CONTINUATION Identify the infrrnint ion thnm is being continued on this shccl ( or cxamplc: Itescript ion or Condition)
Note:
ntrics mndc on this sheet sh:ll he signed nd decd.
I w
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,,,,,t,,,1She.et rnfln ONV 5TU-4.4.43 nev I SITE SAFRIY AsSEsstEnT/EVALUATION OF SSP-27.3 QrAIIfl)ARD
- CIANGES, TESTS, AND Er=ERIMeNTS Rev.
3 r"lCTI CE (10 CFR 50.59)
Page 16 of 18 ArrEnIX D (rage 1 of 1)
PageL1Qfj 3 7-'
.JA-t s..,I DCN! ______~
EHInEERIHG SAFETY SSESSMEfT/EVM.UATIOR COVER SEETJ N
7-n29 n 1 1 e no tpage a3 of I-0 Safety Assessment Only t.y, s n
a n
-d.In t
vl i
X3 Safety Assessment and Safety Evaluation irx it-it RIMS Accession 1lo.
Safety Assessment/Evaluatoioi lo. M62.99B Revision No. -
Rs It llot Rev. 0. PrevlousRevislon's IMS lloB38911129813 Project and affected unit(s) SON Unit 1 Post Accident Sampling Activity Hurnber (Include Revision No.)
Design Change Temporary Alteration Special Test/Experiment Temporary Shielding Request Procedure Change flew Procedure VMP or DG IIo.DCN M-06299B TACF No._
Special Test No.
TSRF No.__
Procedure No._
Procedure Ho._
IR4 Other (Identify)
Special Requirements?
Yes X
Rio Comments:
SPECIAL REQUIREMENTS: See SA section A.3 and DCN M-062 9-B Modification Criteria.
l.
Revision 4 incorporates DCN M-06299-B, final issue.
R4j5'
.0..
Distribution (Safety Evaluation Only)
- RIMS, ET SLE-K I\\-->.
PORC Secretary, DST OA, SQI IISRB Chairman, BR N1 7711-C Site Licensing, OPS AC, SQ11 Site Report tlnit, S13T 4, SQtt Preparer -
Return original to orlginatin documet AA0600n5 I
CG 000028
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I 10 caru 50.59 XVXM.UA=ons or GS, 7ESTS.
S;D r rz2X TTS j STMWAR.D STD-12.13 Rev.
Page 25 of 31 xrmulnix n 5Xra-rX ss XSzMrr rORXXT Page 3. of Document Ito.
(i.e..
CN Ito.,
procedure llo. and revision, special test o., etc.)
D A /Y-Q41 9 9 R-A.
Description
- 1. Detailed description of the change, special activity, or condition Including the ystems, structures, and components affected.
Include the number of the activity proposed (e.g., ECH/DCZ o.,
procedure Ito.).
5eEi AXAC.S\\t)
- 2. References.
AeB ATT AC1iO) la.
Impact on Safety "Is the change. acceptable from a nuclear safety standpoInt7" Yes _
~o No Justificatlont hecklist B-(or IOR procedure change procesasengineoring citerasTD-2.3).
If the answer is Ito', it in an unsafu change which will require either revision to make it safe or cncellation. Environmental impocts are evaluated n accordance with SD-13.3.
C.
Potential Technical Specification (Tir) Impact Yea%
No Ia b change to the T/S required for concticting or mplementing the change, test, or experiment?
i I Justificationt EE A-r TACI-1eD If the answer s Yos", a TS change Is required prior to Implementation or te activity needs to be revised or cancelled, proceed to Part E. If the ansuor to Lhe question is 11o",
proceed to Part D.
CG 000030
-I TVA 40004 tONI'.115:g 7530C AA000137
DCN J C-d-Page -,Q ofl 50.59g zM.Xtt^0lOS r CUmmGS, STD-12.13 rESTS. AND PERIHEIS RV. O
- -n1ell Shoat FoTin 01Il' STD 4.
I I
I
- O CR !
U P
I GSh"XID I
Page 26 Of 31 XrPEl1Di n
Continued)
Document Ito. O 14-1cZ6 2 6
DOES HOT POTEIiTIXLLY DECREASE REDUCES ItUCLtXR 1CLEAR SA.EIY StKEETY 1.
2.
3.
4.
5.
6.
7.
- 8.
9.
10.
11.
12._
13.
14.
15.
16.
.17.
- 18.
19.
- 20.
_l Page I of rire Protection (Appendix )
L-~
Internal Flooding Protection (ELB)
Pipe Breaks Pipe Whlip tModificatlot'to Mon-Selsmic reas In CB/AE ALXRA b-wLJt Impingement ffects SeismIc/Dead elght I-- Internal/External siles
.j:'
fleavy Load Lifts or Safe Load Paths (ITUREG-0612)
Toxic Oases lHazardous Material
-'~
Iluman actors t-~
Electrical Separation/Isolatlon Primary Containment IntegritylIsolation
'-l Secondary Containment Integrity/
Isolation Equipment Reliability M aterials Compatibility
_-Single ailure Criterla Control Room abitability I
'I
- 1 III I
TVA 44 IOflr.1248) 7530C AM300138 TVA 404 10MtP-2.39) 7530C AA000138 CG 000031
DCN1 s 62. Qf l
I Page
- of S;teld"eit Ulwtt I
etn nut, si.t).4.4 3 ite t
10 CR 50.59 r.tmxoNs or CtAGZS, STD-12.13 Br
'TESTS, "D rlR'IMEWl)
Rev. 0 STAUD&RD rage 27 of 31 xrPhiruI D (Continited)
Document to. _DC
^1-Q&Z9FS DOES OT POTEHTIXJL.Y DECRZXSE REDUCES FrUCLZAR lUCLEXR SXFETY sATTry
- 21.
22.
23.
24.
25.
26.
27.
28.
1-_~
Page
- i of hiA Environmental Qualification Category L-~
Equipment Failure odes
_-'Tornado r External lood Protection L-- Protective Coatings Inside Containment Water Spray/Coadensation System Desigo Parameters Test and Reteot Scoping Document (Post Modification Test)
Chemistry Changes or Chemical Release Pathways L.'
Equipment Redundancy L-Equipment Diversity rbysical Separation Electrical Loads L-' Response Time of Emergency Safeguards Equipment
'.. Safety Injection/Core Cooling Capability
~~
Decay loat Removal Capability Reactor Coolant Pressure Boundary t -~~
Reactor Core Parameters Pipe Vibration
_.-"Security System ScaffoldIng 29.
30.
31.
32.
33.
34.
35.
36.
37.t iii 39.
40.
CG 000032 TVA 40O4 tOrl` 1 7S30C AA000139
DCNI.m (.
IZi' I SINI~teid si...,trra,,, OfIl STD-4.4.4
.v Page
,16 10 Cr 50.59 MLLUXTIONS or CnAmaS.
STD-12.l3 H P
- TXStS, AD tX=EZRLS Rev.
Pag 2 of 31 xrrLHxDIt B (Continued)
Document o..
FiQ8ocZ96 Page L of DOES IOT POTEStTIALt.Y DECREASE REDUCES IIUCLEXR I1UCLEAR S.TETY SATETY IT/3
- 41.
Electrical teaker Alignment Changes
- 42.
i I-TABS. Protectlon Relay Settings
- 43.
Compensatory Measure
- 44.
L-Environmental Impact Statement (See STD-13.3.)
- 45.
Design Basis Document
- 46.
_ L-=
Radwaste System Changes
- 47.
Valve Alignment Changes
- 48.
Shield Building Integrity SON/HBlt)
- 49.
e_
2w Radioactive Effluent (Liquid or Gaseous) Release athways
- 50.
5-Temporary Shielding S1.
L Instrument Setpoints
- 52.
'4 SH2 Section XI D.
Potential Safety Analysis Impact Yes to Is a SE obviously required (or eample, change to a Radwaste System. special test, or experlment)?
If yea. the remainder of Part D may be bypassed.
Does the proposed activity affect significantly (directly or ndirectly) any information presented n the SR or deviate from the description given In the SARt Yes ITO 1/A By changing the system design or functional requirements?
Yes It Mo I/A by changing the text, jjsgraphs, or figures?
CG 000033 TVA 4=4 ION-r1-s) 7530C
I~~
10 CR 5 Ur T
STAH~hI1 0.59 EVNM.U1TIOIIS OF CUAHC ESTS.
Pim ExrIME4NTS
- Es#
STD-.A2.13, Rev. a InterIn Change Page 29 o 31 IC M 0799 Page I
M'rrrHDIX 1 (Continued)
Document Hto. ULCN tl-pG299-Page 711 of _
Justification: S AIWNCUEW Does the proposed chnnge involve new procedures or instructions or revisions thereof that:
Yes Ito a/
Il/h Differ with or affect systemi operation characteristics; from that described n the SR7 Differ with or affect compliance w'ith Technical Specifications?
Yes Ho
/
II/?'
Yes Ito/
II/A
.Conflict with or affect a process or procedure outlined# summarized or described in the SR7 Justification:
Se C j-a-ic.Q.
If he questions are answered "Ito",
te activity may be mplemented without further evaluation.
If the questions are answered Yes-, a 10 CR 50.59 S is required.
E.
Review and Approvals I
Preparer (Level 1) -D
(~
A/¶ s,
/
(?s Name Sigr
/.flow &
iature 1Date : e*2.dt1ZL..
Da te:______ 9 __
Reviewer (Level I-IOR tM~.P~.LL
/ 7 or Level 1I)
R Ilame S ntu e Approver (Line gr.)
/JAVsip
/ tli&j
/
F!
e MA~S Date 1/K/ill Ilame Signature v1.'i 7.4EC,1,j.
Other Reviewers (as appropriate)
/
Date:______
Ilame/organization Signatitre 7880oC TVA aol
'os 'C)N th as, CG 000034 AA000I1
stsu;sissa siset row,,% oiit, sro 4.4 1tev liP
,STXllDXRD I
2n CFR 5n.59 EVALUATOI11S OF CnAxicS.,
TsrhSTS. AIm Exrr.UIFz.lrs STD-12.11 Rev.
Page 30 )f 31 I
ArrEnlnix c SAFTYm EVAI.UNTIO11 ORMXT lDCNI M-O 06 z9s B Page 15 of..,
Page of Satety Evaluation 10.:
DdA/
A4 -.QZ 7
A Document lo. (that s,
.CIT lo., procedure lo. and revislon.
special test lo., etc.): -cA iZ A. Accidents valuated as t Design asis X.I. Design asis ccidents nd Anticipated Operational Transients SL=
A Ac.AET)
A.2.
Credible failure modes of proposed activity eEr A-17AcAV-l!
- 13. Evaluation of Effects B.1.
Hay the proposed activity ncrease the probability of an accident previously evaluated n tho SAR?
Yes l
Justification:
SEE A7rr7AC4EO 13.2.
ay the proposed activity increase malfunction of equIpment mportant the S7 S the probability of occurrence of a to safety previously evaluated n
Yes Ito L-Justification: Se A 7rAcXWT 11.3.
Hay te proposed activity increase previously evaluated In the SAR?
thre consequences of an accident Yes Ito Justifycation: 56 A-r-Ae4JC6 B.4.
Hay the proposed activity ncrease the consequences of a malfunction of equipment important to safety previously evaluated n the SARt Yes 110 Justitication : 6 A 7r/CH'iL D.5.
Hay the proposed activity create a possibility [or an accident o a different type tnn any evaluated previously In the SAR7 VA It jci r.:tl'-12.3 1 1 7530C CG 000035 AA000142
- ,I@i*1tA.6 Sleiet Vntm MISP S 1
.... 4 3 Ie I
R t
ST"lDABRD 10 CrR SO.S9 zVALUxTIO1S 01 C
G, 7-rs-z, A
rXR1"IHES S-M-12.13 Rov.
o Pag 31 of 31 I
VDL f -I I -
xrrEirDix C (Continued)
Page.LUb-l ShrmEr EvKLUXUIOR FORH4LI Safety Evaluation l.:
OCA/
M- 062 g 89 Page.
of Yes Il _
Justificationt 7C1Iacac.
S.6.
Hay the proposed activity create a possibility for a malfunction of a different type than any evaluated previously n the SR7 Yes o_
l tV Justification:
Se--
/J7c-xAc.cQ B.7.
May the proposed activity rduce the margin of safety as defined n the basis for any Tchnical Specification?
Yes _to J
Justificationt 5e.c.
qN1r AFLcQ C. UnrevIewed Safety uestion Determination Conclusion Based on the results of Part B, the change:
J_____
Doea not nvolve an unrevlewea safety question.
Involves an :inreviewed safety question and must be revised, cancelled, or reviewed by the 11RC prior to mplementation.
n. Reviews and pprovals Preparer (Level r
,_1-A..
Date t Ilame Signature Approver (Line Mgr.)JJ'/A 54R
/
Z a
Zt AS Date:
Z/t1 g 5 1ame
, Signature V 7.zF..,Ra/
Reviewer (Level I)
Ta-M s C. PO erf lName
-. a..
Signadure Reviewer a,b)
(PORC/QR)
Itame Signature Other Reviewer(s)
/
(t\\ appropriate) llama Signature a his review 1 not required for corporate level procedures.
b s required by Technical Specification.
Date:_l_2__
Date:
Date:
TVA 4I IONr.2-88) 730C AAO001 4 3 CG 00036
LiOLNI
_______I Page 1
of SAFETY ASSESSMENT U
DCN -06299D A.
DESCRIPTION
- 1.
Design Change Request (DCR)
No.
3250 (Ref. A.2.a) addresses various modif ications to the Sequoyah Unit 1 Post Accident Sampling System (PASS).
DCN M06299B (Ref.
A.2.j) implements one of these modifications to install 1/8" OD stainless steel tubing, Tefzel tubing, and a plug type isolation valve withlR3 all necessary f ittings and supports to directly connect PASS lines CP-L-11 and CP-L-12, bypassing the 4-way valve RC-DV-
- 2. (See the attached Figure 1.)
Note that an existing penetration in the PASS panel will be used for the handle extension for the plug valve.
As presently configured in the Ul PASS, the line CP-L-12 is 1/8" OD Teflon tubing.
Line CP-L-l1 is 1/4 "
OD stainless steel tubing.
ine CP-L-12 would see only diluted RCS off-gas samples; however, after the modification stipulated by this DCN (06299B),
CP-L-12 would see both diluted and undiluted samples.
For this reason, this modification will also replace CP-L-12's Teflon tubing to Tefzel tubing, which]3 is capable of withstanding a much higher integrated radiationjRq dose.
This modification will allow the PASS panel operator to directly obtain a sample of RCS off-gas which has not been diluted.
This is necessary so that backup, grab sample capability for dissolved hydrogen concentration analysis is available as required by UREG 737 (Ref.A.2.m) and Reg Guide 1.97 (Ref.A. 2.n).
Prior to the modification, the off-gas sample was taken through valve RC-DV-2 which diluted the sample by a factor of approximately 1:15,000 (Ref.A.2.k).
The hydrogen gas concentration in the sample would be diluted to the point of undetectability by existing radiochemistry laboratory methods (i.e.,
gas chromatography).
With this modification, the off-gas sample receives essentially noi 4 dilution (Ref A.2.e),
and thus the hydrogen concentration remains at levels which are detectable (ef.A.2.p).
This modification allows the acquisition of an undiluted sample during accident and transportation to the radiochemistry laboratory.
The scope of this Safety RI Assessment/Evaluation is (1) the modification itself, (2) the drawing of the sample and placement of the sample into the transport cart/cask and (3) the transport of the sample.
Systems Affected:
Sampling System (System 43)
PASF IIVAC (System 30)
CG 000037 AA000144
IPage L
SAFETY ASSESSMENT 11 DCN
-06299B Components Affected:
Post Accident Sample System Liquid Sample Panel (PASS LSP)
Affected Structures:
None implementing te proposed activities will require revisions to and deletion of existing plant procedures, and/or the issuance of new plant procedures of the following types to maintain and operate the equipment during normal plant conditions:
Maintenance Instructions (MIs)
Instrument Maintenance Instructions (IMIs)
Surveillance Instructions (SIs)
System Operating Instructions (SOIs)
General Operating Instructions (GOIs)
Special Maintenance Instructions (SMIs)
Periodic Instructions (PIs)
Technical Instructions (TIs)
These procedures shall integrate the requirements contained within this design change package.
Revision to these procedures for reasons other than the implementation of this modification are not within the scope of this Safety Assessment/Evaluation.
AA000145 CG 000038
DCNI Moca'i'i6 Page 19 SAFETY ASSESSHENT P
J DCN
-06299D 2,
References:
- a.
Sequoyah Design Change Request (DCR); No. 3250.
- b.
Design Input Record (DIR) 11-DI-574-01, Gilbert/Commonwealth, Inc., rev. 0.
- c.
Vibration Checklist, RIMS# B39 91 0422 002.
- d.
General Design criteria
- 1)
GDC SQlN-DC-V-9.3, rev.
3
- 2)
GDC SQ11-DC-V-3.0, rev.
6
- e.
"Reactor Coolant Off-Gas Sample Dose Evaluation" SQN-SQS2-0120, rev. 1.
- f.
UFSAR Sections:
- 1) 1.2.2.11 -
Sampling & Water Quality System
- 2) 7.5 Safety-Related Display Instrumentation
- 3) 9.3.2 Process Sampling System
- 4) 9.4.10 Postaccident Sampling Ventilation System
- 5) 9.5.10
- Post Accident Sampling Facility
- 6) 15.5.3 Environmental Consequences of a Postulated Loss of Coolant Accident
- g.
Technical Specification Sections:
f:
- 1) 1.0 Definitions
- 2) 3.3.3.7 Post Accident Monitoring
- 2) 6.8.5.e Postaccident Sampling
- 3)
Licensing Condition 2.F-Post Accident Sampling
- h.
Project Planning Document (PPD) No. 0608; February 7, 1991.
- i.
TI-66.2:
"Sentry Post-Accident Sampling Equipment Training and Test Results Comparison",
rev. 7.
- j.
DCN M06299B IRA
- k.
TI-66.1:
"Post-Accident Sampling and Analysis Methods for the Sentry Post-Accident Sampling System:, rev. 12.
- 1.
TVA Drawings, 471625 series
- m.
HUREG 0737, "Clarification of TMI Action Plans:
- n.
Reg Guide 1.97, "Instrumentation for Light-Water-Cooled Nuclear Power Plants to Access Plant and Environs Conditions During and Following and Accident".
- o.
"Safety Evaluation by the Office of Nuclear Reactor Regulation Related to Amendment No. 35 to Facility Operating License DPR-77 and Amendment No.
26 to Facility Operating License DPR-79 Tennessee Valley Authority"; TVA RIMS! A02 840426004.
- p.
Sensitivity Assessment in DCN M06299B.
- q.
SQEP 26 att. 12 checklists in DCH M06299B.
- r.
ALARA Review in DCN M06299B and Mechanical Nuclear 4
Instruction MNI 25.5.1.
- s.
Bill of Materials in DCN M06299B.
CG 000039 AA0OO.4Sj
DChl
-'Z 7 913 Page 2Z of 13 SAFETY ASSESSMENT DCH M-06299B t.
U.
V.
w.
x.
Y.
z.
aa.
bb.
cc.
dd.
ee.
ff.
gg.
fill.
"H1ligh Radiation Sampling System Operating and Maintenance Manual"; Sentry Equipment company, TVA Contracti 80X62-827371.
Calculation for total PAS mission dose, TI-RPS-146, rev. 0.
SQN T/S Change equest# 89-30, PAM Instrumentation, TVA RIMS# L44 910412 801.
TVA Letter to IJRC on Reg Guide 1.97 deviations, TVA RIMS# L44 900507 804.
CRFSAR (copy attached) calculation "Seismic Qualification of Liquid Sample Panel 1 & 2-L-567"; SCG-4M-00676, rev. 4.
CAQR#SQP 90 0191 (Generic review located mission dose problems.)
Memo: Vondra to Trudel, TVARIMS1 S52 90 0907 860.
Design Input Memorandum (DIM), SQN-DC-V-9.3-10.
SSP 27.3, "Safety Assessment/Evaluation of Changes Tests and Experiments (CFR 50.59)" Revision 3.
Calculation TI-RPS-146, Revision 4, "Post AccidentlRA Design Basis Sampling Mission Dose:.
SQA-66, "Housekeeping", Revision 18.
RCS Off Gas CART, Drawing
- o. 464, Atlan-Tech, Inc.,
Purchase Requisition No. 194713.
IZSAC-125, Section 3.6.
Si-160, Revision, 4, Primary Containment Local Lakrate Summary.
ii.
Calculation SQN-APS3-067, Revision 1.
jj.
Calculation -
Tubing Material Qualification of off gas sampling line, SQN-43-D05j, EPM-MDE-120491, revision 0.
kk.
Calculation - Integrated Dose for off gas sample tubing, SQN-43-D053, EPM-CSS-120291, revision 0.
- 11.
Calculation - Integrated Accident Dose inside of Primary Rt Containment and the Annulus, TIRPS48, revision 5.
mm.
Memo -
Sentry Corporation (William Wagner) to TVA (SWEC)
Bob Weth, dated 12-02-91, (B27911202003)
- 3.
Special Requirements:
- a.
This SR deleted in Revision 2.
- b.
Calculation "Reactor Coolant Off-Gas Sample Dose Evaluation:
(Ref A.2.e) contains an Unverified Assumption (UVA). This UVA says "The 30 second handling time specified by Chemistry includes time for drawing the sample and placing it in the sample carrier.
Thisl
.duration must be verified after the modification islt performed." This must be done prior to lifting the hold order on valve RC-V-23. The UVA must be resolved while the hold order is on valve RC-V-23 by utilizing a "dry run" approach. This UVA was verified per revision 1 of Reference A.2.e.
RA AA000147 CG 000040
page 2
SAFETY ASSESSMENT DCN H-06299B 14
- c.
This Safety Assessment/Evaluation is written assuming only post-accident and training exercise usage of the modification.
Any procedure initiation or revision forj 2 normal sampling will be addressed by DCN M06550.
- d.
Verification of tubing, valve, and support installation requires E-CEB approval prior to declaring the system R4 operable.
For. verification, see sheet 32 attached.
- e.
This SR deleted in Revision 1.
- f.
Attachment B to the Modification Criteria provides guidelines for installation of the stainless steel tubing. Deviations from these guidelines require NE-CEB approval via F-DCU.. Maximum segment lengths (see Ref j Mod Criteria) are not to exceed 25 inches.
- g.
This SA/SE addresses only Unit 1. Therefore, one of the UVAs contained in reference A.2.dd does not affect this SA/SE, but does affect the Unit 2 PASF mission.
However, the operator duties as assumed in Reference A.2.dd must be implemented in the required procedures (i.e.
Ti-66, Reference A.2.i and A.2.k, etc.)
and reviewed by RADCON.(see ALARA Review).
I h.
The other UVA in Reference A.2.dd concerns the sample transport cart.
The sample transfer cart must be verified to reflect the recommendations of Re Jerence A.2.dd (TI-ROS-146). This UVA was verified per r ision 4 of reference A.2.dd.
I R2
- i.
The REP-IP for operation/staffing of the Osc must be reviewed and revised, if needed, to include the staffing R5 requirements of reference A.2.dd (the manpower requirements for RCL analyst and RADCON techs.)
B.
Impact on Safety B.3 Pipe Break:
Tubing is TVA Class G or equal, non-safety I4 related (Ref.
A.2.b and A.2.jj)
Technical Evaluationl Checklist SQEP-26, rev. 9, Att. 12) performance for Pipe Rupture resulted in complete "No" answers for this modification (Ref. A.2.q).
Therefore, pipe (tubing) break for this modification is not credible and is not a concern for plant safety.
B.6 ALARA:
An ALARA review has been completed per SQH SQEP-26 2 and Mechanical/Nuclear Instruction M/HI 25.5.1 (Ref. A.2.r).
uclear safety is not compromised due to ALARA considerations. A revised ALARA review was completed for the sample transport.
AAooo4S CG 000041
DCNI Y o0z-99B Page z-of SAFETY ASSESSMENT 15 DCN M-06299B B.8 Seismic/Dead Weight:
All tubing and fittings that will be added via this modification will be installed per Attachment A and B of te odification riteria.
To ensure the tubing Ri is qualified o seismic category I(L)B for QR related TVA Class G portion of the system and to ensure the seismic qualification of te sampling panel is not compromised, Calculation SCG-4M-00676 was performed (ef.A.2.y).
An existing spare penetration will be used for the valve stem extension.
This precludes the need of creating a new penetration.
This will therefore not affect the seismic design of the panel structure.
The RCS off gas transport cart has locking front casters (reference A.2.ff). Per SQA-66 (reference A.2.ee) these willj be locked while the cart is unattended to prevent interaction RZ with equipment important to safety.
Thus, no adverse effects are created by the seismic/dead weight considerations of this modification.
B.ll Toxic Gases: The RCS off-gas sample is a small volume (15cc) and does not contain toxic gases, only radioactive gases.
The operators performing the sampling process shall be equipped in self-contained breathing apparatus for radiological considerations. Additionally, the Post-Accident 1IVAC System, the ABGTS, and the RCL exhaust loads shall be in service to mitigate the consequences of airborne contamination. Therefore, there are no adverse effects upon Rz nuclear safety due to toxic gases considerations.
B.15 Primary Containment Inte rity/Isolation: When the CIe s are opened, the plant will enter T/S L.C.O 3.6.1.1, Containment Integrity, which has a one (1) hour action time.
The time R constraints of L.C.O. 3.6.1.1 and the ability to close the sample CIV's make this not a concern for nuclear safety.
B.17 Equipment reliability:
The proposed bypass line will be a modification to an existing vendor supplied system.
The modification will add a sampling feature to the system, but existing features and equipment will be unaffected. No other change in the way this system operates occurs.
Nuclear safety is not compromised form the standpoint of equipment reliability.
AA00014 9 CG 000042
DCNI D Laid-Page
-L of SAFETY ASSESSMENT ari DCN M-06299B
,B.18 Materials Compatibilitv: Te additional fittings, and valve are specified to be stainless steel.
The additional tubing is specified to be stainless steel and Tefzel (Ref. A.2.s).
This is identical to the materials now in use in the system with the exception of the Tefzel tubing being used in lieu of the vendor supplied Teflon (Ref A.2.t).
The Tefzel tubing R4 will withstand te increased radiation exposure caused by the undiluted sample and is more than adequate for the design pressure and temperature (Ref A.2.jj and kk). The PASS panel supplier, Sentry Equipment Corporation has concurred with the use of Tefzel as a suitable replacement for Teflon (Ref A.2.mm).
Thus material compatibility is not to nuclear safety concern.
B.20 Control Room habitability: No effluent release pathways are changed nor are postulated releases noticeably increased by this modification; PASF VAC will function unchanged (Ref.
A.2.f.4).
The ABGTS will function during sample transport.
In order to analyze samples in the RCL, the RCL support systems (detectors, lighting, VAC) must be available.
The contribution to outside air contamination (the release "plume") from the PASF IIVAC will not change due to this modification (Ref. A.2.f.4).
the Control Room pressurizing Ra fans will draw outside air during an accident, including the release from the PASF IVAC/ABGTS/and RCL exhaust via the service building vents.
Since the release from these pathways has not noticeable changed (see item B.53), dose to the operators in the Control Room has not changed.
Therefore, this modification has no effect on Control Room habitability.
B.21 Environmental Qualification categorv:
The modification itself does not have any effect upon Environmentally Qualified equipment (Ref SQEP-26, Att. 12 checklist for DCN M06299B). After the modification is in place, no incrbase in total integrated dose for the PASF room will occur duelto the modification not increasing the amount of radioactive material in te PASF room (Ref A.2.f.4).
The dose from the transfer cart to EQ equipment is insignificant (reference A.2.dd).
Thus, nuclear safety due to EQ categories is not a Z concern.
AAO0015 0 CG 000043
DCNI
-OLLI 5_
IPage 2-. of
-g SAFETY ASSESSMENT 19 K)
DCN
-06299B
,other)
B.53 GDC 19. Offsite Dose. Access To Vital Areas:GDC 19 limits dose to plant personnel to 5 rem whole body following an accident.
The operator dose due to the collections, transport, analysis of the undiluted Rcs off gas sample has been calculated by reference A.2.dd to be within GDC 19 limits.
Offsite dose will remain tie same since te quantity of radioactive material released from the PAST TIVAC does not increase due to the modification (Ref A.2.f.4).
The offsite dose contribution due to a direct release (RCL) from the minuscule 15 cc vial is negligible when compared to the volume released from containment due to assumed containment leakage post accident (an instantaneous 15 cc release compared to La = 225.17 SCFII (Reference A.2.1hh) or 6,376,011 cc/hr).
The RCS volume is 12,612 ft3 and the containment volume is 1.2 x 10' ft' (Ref.
A.2.f(6)). Given that the off-gas sample of RCS will be identical qualitatively to the containment atmosphere, the quantitative amount of curies (e.e.,
the concentration in Ci/cc) in the 15cc sample vial will be approximately 100 time greater than the Ci/cc of containment atmosphere based solely on the volume differences between RCS versus containment.
(The containment has a greater volume for dilution purposes).
The containment
- leakage, L,,
is 6.37 x
106 cc/lr (1.53 x 10' cc/day, Ref. A.2.hh).
Thus, even given the 15cc sample vial can be assumed to be equivalent to 15 times 100 or 1500 cc of containment atmosphere, it represents 1500 divided by 1.53 x 10, or less than 0.001%
of the total expected release.
Further, te 15 cc sample vial will contain largely noble gases. Whereas the release from containment above usually will be filtered, the 15cc vial will not be.
however, filtering is not an effective means of impeding noble gase release and thus the lack of filtering prior to release is not an issue in this case.
The volume leaked from containment and assumed in the design bases to go to the atmosphere unfiltered 25 of containment leakage goes to the AB and AGTS is 4elayed 5 min. and 75% goes to annulus and EGTS delay is 30 seconds (reference A.2.ii and A.2.f (6)) is:
AB(unfiltered)=0.25 (6,376,011 cc/hr) (lhr/60min)(5min)
= 132,834cc Annulus(unfiltered)=o.75(6,376,oll cc/hr) (lhr/3600s) (30s)= 39.850cc
,'tal unfiltered
= 172,684cc AA000151 CG 000044
SAFETY ASSESSMENT DCN M-06299B DCNI -m-o.i_
Page 2
of l
_~~~21 Comparing te equivalent unfiltered volume of 1500 cc to the above 172,684 cc shows an increase of less than 1%i on the total amount of unfiltered leakage going directly to the environs.
It should also be acknowledged (as stated in FSAR 15.5.3) that the highest activity assumed inside primary containment (Reg. Guide 1.4 LOCK) cannot occur coincidentally with the highest activity assumed for a PAS sample (i.e., the RCS is assumed intact and pressurized in the PAS dose analysis).
Therefore, the 15cc sample vial being taken outside the ABSCE is a negligible contribution to 10 CFR 100 or GDC 19 doses and is not a concern for nuclear safety.
During transport of the sample, it will be contained within a shielded transport cart; thereby minimizing operator dose and the dose contribution to the general area (3.5 mrem/hr at 18" per page 12.30 of reference A.2.dd).
Thus offsite dose and vital area access are not a concern for nuclear safety.
'other)B.54 (other) B. 55 Explosive Gases: The sample vial will be drawing a 15 cc sample of RCS off-gas.
The concentration of hydrogen gas in the sample will be quite low, and the size of the sample (15cc) is very small.
Explosive tendencies are quite unlikely.
Therefore, explosive gases are not a\\
concern for nuclear safety.
PASF Manpower: Reference A.2.dd requires at least 5 RCLI analyst in order to complete the design basis (first) mission to the PASF and subsequent analysis in the RCL.
Reference A.2.m requires daily sampling for 7 days and once a week till 30 days after the accident. Reference RZ A.2.dd calculates RCL analysis dose at close to the GDC 19 limits for the first mission. Therefere, 5 Dr 7 -
5 R-r-analyst-are-equ-ed-and-s ary nt -
14 tt RA9GON-teehs-are-equired--the firstS-week.
The-a b4li-ke-suppert-th s-PAS-4unet-ien-and-eenti-nue e-- sper-t the Rq areund-the-eeek-shi-fbt une-tien-4n--the Chem-T4,ab--and RA9DGNI-
-i n-epen--4tem--that--m-s'-b eb-spee s
eearly addressed-.
The Radiological Emergency Plan ('(REP)
Implementing Procedures (IP) for the*
OSC R2.
operation/staff ing must be reviewed and revised if needed to address this item before utilization of the R PAS to obtain undiluted RCS off gas samples (see SR R A.3.i).
AAOQOL5 2 CG 000045
I-CNI -"-
cac-2ci6t I Page zB of SAFETY SSESSHENT 21 DCN M-06299D It should be noted that according to FSAR Figure 15.5.3-7, the dose rate due to contained sources throughout the RZ plant decrease substantially over the 30 day post-accident period and will result in decreased operator dose rates.
In fact, computer output used in TIRPS-48, (Ref. 2.A.11, Ilicrofiche TVA-F-G100850: Executed Nov.
19, 1991) indicates the gamma shine inside containment is reduced approximately by a factor of ten, 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 4
following the large break LOCA.
This information is consistent with FSAR Figure 15.5.3-7.
Additiena--m-pewer an-be-breugh tin -rem-FNWBN-fer PAS--train and-dut4es-Section B Conclusion By the reasons given, this modification causes no adverse impact on nuclear safety.
C.
Potential Technical Specification (T/S) Impact (Cont.)
Justification: T/S 6.8.5.e and Licensing Condition 2F address Post Accident Sampling.
2 T/S 6.8.5.e requires the plant to have the capability to obtain and analyze various samples under accident conditions. This modification does not affect the PASS's ability to perform its original design function:
diluted RCS off-gas samples will still be capable of being taken.
The modification will allow undiluted RCS off-gas samples to also be taken.
The licensing Condition 2.F addresses the requirement for having a Post Accident Sampling System, but this modification does not affect, remove or change the Pass except as stated above.
tlothing in this modification affects either TS 6.8.5.e or Licensing Condition 2.F.
T/S 3.3.3.7 for Post Accident Monitoring (PAM) requires Type A and/or Category 1 PAM variables to be addressed.
Since the PASS is Category 3, this T/S is not Affected.
(Ref A.2.v)
T/S indirectly related to this modificatiok are:
- 1)
This item deleted by Revision 1.
lR?.
K>
AA000153 CG 000046
PDCge 0-Page -K f-SAFETY ASSESSMENT DCN H-06299D
- 2) 3.6.1.1 -
Primary Containment Integrity
- 3) 3.7.7 -
Control Room Emergency Ventilation
- 4) 3.7.8 -
Aux Bldg Gas Treatment System None of the above indirectly related T/S items are affected by the modification.
D.
Potential Safety Analysis Impact (Cont)
Justification:
The modification does alter information presented in the UFSAR.
No UFSAR text, graphs, or figures are involved; however, Table 7.5-2 will have an addition to state that the PASS has the capability for RCS off-gas grab sample acquisition for dissolved 112 analysis as stipulated in Deviation 22 to Reg Guide 1.97 (Ref A.2.w).
No change to system design, function or operation as described in the UFSAR occurs.
The modification does not change a radwaste system. The modification is not a special test or experiment.
Full compliance to T/S is obtained as described in Section C of this Safety Assessment.
Placing the bypass line and plug valve around valve RC-DV-2 does not adversely affect the existing sampling system.
Procedure changes must be screened per the criteria in Site Standard Practice SSP-27.3.
Since the modification involves a USAR change, a safety evaluation is required.
A CRFSAR has been prepared (Ref A.2.x).
The Safety Evaluation also addresses FSAR sections 9.4.10 ad 15.5.3 (below) with respect to (1) offsite dose, (2) operator dose (both in the control room and in the PASF) and (3) access to plant vital areas.
UFSAR Sections Researched:
1.1.2.2.11
- 2.
7.4
- 3.
9.3.2
- 4.
9.4.10
- 5.
9.5.10
- 6.
15.5.3 Sampling & Water Quality System.
Safety-Related Display Instrumentation.
Process Sampling System.
Post Accident Sampling Ventilation System.
Post Accident Sampling Facility.
Environmental Consequences of a Postulated Loss of Coolant Accident.
K>
CG 000047 AAOOO54
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(C Senntrg Equipment Corporation High Radiation Sampling Sgstem Reactor Coolant Sampling Flow Diagram Diluted RCS Off-Gu Sample Vi,
.kr-on In (100 plig)
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DVI V8,1 V16 RCS Duted Sample Redrawn rom SQNP 1k66.1. Rev. 12.
Appendix B, p. 10 of 7
.,AI M-OLL956 page 3 of ShFETY EVALUhTIO0lZ P
DCU M-06299B A.
Accidents Evaluated as the Design Basis A.1 Design Basis Accidents and Anticipated Operational Transients:
UFSAR 15.5.3 Events:
- a.
Environmental Consequences Due to the Operation of the Post-accident Sampling Facility.
- b.
Plant Accessibility Post LOCA.
A.2 Credible Failure Modes of Proposed Activity:
- a.
Transport athwav events -
getting the sample from the PASF to the radiochemistry laboratory.
- b.
Valve failures -
i)
Fail closed Undiluted sample can not b'e obtained.
ii)
Fail open/leak Affects the acquisition of diluted RCS off-gas samples.
- c.
TubingrLupture the tubing added by the modification 2
experiences rupture.
- d.
Dropping of sample vial -
i)
Vial breaks.
ii)
Vial does not break.
B.
Evaluation of Effects.
B.1 Proposed modification will have no effect on the probability of occurrence of any accident previously evaluated in the UFSAR. The modification is located entirely within a vendor supplied system which is categorized TVA Class G piping or equivalent, non-safety related, and the modification itself is also Class C or equivalent (see Modification Criteria for R4 DCN M06299B). In an accident case, the modification will not be used except when an accident (LOCA) has occurred.
The probability of an accident occurring is not increased because the accident has already occurred for this system to be utilized.
CG 000049 AAOQOO(!
Page 3 of l SAFETY EVALUATION 25 DCN -06299D In a training situation, T/S LCO 3.6.1.1, Containment Integrity must be entered as specified.
Performance of sampling techniques will e identical to those prior to the modification.
The only exception being the acquisition of a quantity of off-gas while the source of this off-gas has been isolated from the RCS.
Entry into LCO 3.6.1.1 with its restrictive 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> Action time minimizes the probability of an accident while containment integrity is not established.
Therefore, there is no increase in the probability of occurrence of any accident previously evaluated in the UFSAR.
B.2 The proposed modification will not increase the probability of occurrence of a malfunction of equipment important to safety.
The modification and all equipment associated with it are TVA Class G
or equal, non-safety related.
The Tefzell tubing which replaces the vendor supplied Teflon has been qualified to its design temperature,
- pressure, and 4
applicable radiation exposure (Ref 2.A.jj and kk).
Vendor concurrence was obtained from Sentry Equipment to substitute Teflon with Tefzell (Ref 2.A.mm).
Any equipment important to safety in the area is not adversely affected due to this modification because the equipment will not see an increase in radiation dose since the modification does not change the t_ total integrated dose to the PASF.
Similarly, no increase in
\\.temperature, pressure and humidity are expected because the modification does not alter existing lines which could have environmental effects on equipment important to safety.
During the sample transport from the PASF to the RCL, the 15 cc vial will be enclosed in a shield cart (pip),
it will be attended by at least two individuals (reference A.2.dd and SR A.3.g) and has locking front casters to avoid interactions with equipment important to safety if left unattended (reference A.2. ee).
Thus no impact to equipment important to safety is seen, and therefore, there is no increase in the probability of occurrence of a malfunction of equipment important to safety.
CG 000050
N M-oa99J3I IPage 3-_of SAFETY EVALUATION DCN -06299B B.3 The proposed modification will not increase the consequences of any accident previously evaluated in the UFSAR.
The environmental consequences due to the operation of the PASS discussed in UFSAR Section 15.5.3 are based on the volume of reactor coolant accumulated in the post-accident sample collector tank (SCT) which is vented to the PASS 1IVAC System.
The credible failure modes result in release of activity to the PASF atmosphere (outside the sample panels where the operators will be performing the sampling procedure) which is also vented to the PASS 1IVAC System.
The credible failure modes do not increase the amount of activity postulated to be released to the exhaust system, only the pathway by which it reaches the PASS IIVAC System.
In other words, the quantity of radionuclides does not change. Instead of being deposited directly to the PASF IIVAC filters as UFSAR 15.5.3 describes, the credible failure modes will first put a small fraction of airborne gas into the room instead of to the SCT.
From the room the PASF 11VAC will remove the airborne gas and filter it through its filter housings. The pathway is changed, not the end amount of radionuclides or the quantity or the final location (the PASF IVAC filters).
Therefore, the offsite
. dose due to this change in operation of the PASS is not increased.
Similarly, PASF 1IVAC will continue to function unchanged.
The contribution to outside air contamination (the release "plume") from the PASF 1IVAC will not change due to this modification (Ref. A.2.f.4).
The Control Room pressurizing fans will draw outside air during an accident, including the release from the PASF 1IVAC.
jR?
Sampling operator dose is calculated in part by Ref. A.2.e of the Safety Assessment. The total mission dose calculated in reference A.2.dd is within GDC 19 limits provided the operator duties are distributed as assumed (see SR A.3.g).
During the sampling transport from the PAS to the RCL, the ABGTS will be in operation to holdup and filter any release.
During transport, the sample is contained in a shielded (pig) cart, attended by two individuals, and secured from moving RZ (wheels locked) if left unattended.
This provides adequate control to guard against a release from the vial during transport.
While inside the CL, necessary support functions (lighting, power supplies, detectors, JIVAC) are assumed available or there would be no basis for returning a sample there for analysis.
Any release from the RCL is via the exhaust hoods and their IEPA filters (no charcoal) to the service building (SB) vent (a monitored point).
M0OOO 56 CG 000051
r oi vA-o.zFfi%
Page of _
SAFETY EVALUATION 27 DCV M-06299B The worst case release would be the discharge of the entire contents of 15 cc vial directly to the CL and out an exhaust hood to the SB vent taking no credit for the non-ESF filters.
The release (15cc), is diluted within the RCL and exhausted via an RCL hood, further diluted by other flows in the SB vent and finally discharged to the environs.
This 15 cc release is insignificant when compared to the existing containment leakage.
See SA item B.53 for more analysis.
Z Therefore, the doses resulting from operation of the PASF are within the 10CFRloo limits.
As such the changes in consequences are so small it cannot be reasonably concluded that the consequences have
- changed, thus the transport/analysis of the sample need not be considered an increase in consequences (Reference A.2.gg, page 3-5 and 3-6).
') Access to plant vital areas is. not affected since the PASF is located in an enclosed room in which no equipment required for safe shutdown of the plant or to mitigate the consequences of an accident is located and to which access must be available.
The transport cart is sufficiently shielded to reduce the general area radiation dose such that Rt access/egress is unaffected.
B.4 The proposed modification will not increase the consequences of a malfunction of equipment important to safety. As stated before, the modification and all equipment related to it (e.g., the PASS) are non-safety related. The only equipment important to safety associated with the PASF in the area of the modification are te containment isolation valves.
Addition of the modification will not cause an increase in the dose rate or other parameters as explained in Part B.2.
As such, the proposed modification will not increase the consequences of a malfunction of equipment important to safety.
B.5 The proposed modification will not create the possibility for an accident of a different type than evaluated in the UFSAR.
The modification is isolable from the RcS by containment isolation valves and therefore can not adversely affect the RCS to create an accident of a different type.
In an accident case, if te modification is in use, then an accident has already happened.
Therefore, the accident is already in progress and was not created by the modification.
No other credible accident scenario could be postulated.
AAO0 0 1 5 9 CG 000052
Page -3, SAFETY EVALUATION 28 DCN M-06299B In a training case, since the CIVs are opened for training purposed by approved plant procedure TI-G6.2 (Ref. A2.e),
LCO 3.6.1.1 is applicable and shall be addressed.
The time constraints of the hour Action time for LCO 3.6.1.I and the ability to close te CIVs preclude the creation of an accident of a different type.
- Thus, the proposed modification does not create the possibility for an accident of a different type than evaluated in the USAR.
B.6 The proposed modification will not create the possibility for a malfunction of a different type than evaluated in the UFSAR. The credible failure modes listed in part.2 will be addressed separately:
- a.
Transport athway events The first radiological assessment (Ref A.2.e) has as its scope the acquisition of the sample and placement of the sample in the sample carrier.
Another radiological assessment (reference Z
A.2.dd) was performed for the mission dose.
Special requirements A3.b, A.3.g, and A3.h must be satisfied to ensure the GDC 19 dose limits are met.
- b.
Valve failures i)
Fail closed Undiluted sample can not be obtained.
However, a diluted sample can still be obtained as currently available.
ii)
Fail open/leak Affects the acquisition of diluted RCS off-gas samples.
Sample line can be isolated using existing valves to terminate the release to the PASF (see Figure 1, valves RC-V-l1 and RC-V-9, as well as containment isolation valves).,
SCBA is in use by the panel operators to provide protection from airborne contamination (accident situation only).
PASS 1VAC is in service to mitigate the airborne release (Ref A.2.f.4).
Since the valve failures can be mitigated using existing equipment, the modification does not create the possibility for a malfunction of a different type than described in the UFSAR for valve failure.
AA000160 CG 000053
Page 31 of.
SAFETY EVALUATION 29 DCN
-06299D C.
Tubinq ruptxure -
All tubing is located inside the sample panel.
Undiluted RcS off-gas lines (constructed of stainless steel) already exist inside the panel before the modification. The existing Teflon tubing (CP-L-12) is being replaced by Tefzell tubing by thisItR3 modification, and the new bypass line is made of stainless steel.
Vendor concurrence to use Tefzell in lieu of Teflon was obtained (Ref 2.A.mm).
The Tefzell tubing has been qualified for design pressure, R4 temperature, and applicable radiation exposure (Ref 2.A.jj and kk).
Therefore this is not a new credible failure.
- d.
Dropping of sample vial -
i)
In an accident situation, an unbroken vial affects operator dose by increasing exposure time while retrieving the vial and placing it in the sample carrier.
The radiological assessment for DCcN M106299B assumes 30 seconds to get the sample into the vial and the vial into the carrier with total dose in this 30 seconds of 75 m whole body (worstIR.
case).
If the vial can be retrieved within the 30 second time-frame, then retrieval with tongs shall be performed.
If the vial orients itself in such a manner that retrieval is not possible, then the sampling operators shall vacate the area to assess the situation.
RADCON and the Technical Support Centers are operational to provide guidance.
For training cases, if the vial is dropped and does not break, it will simply be retrieved, and the exercise can continue.
Once the training sample has been obtained (i.e., the 15 cc vial is filled), the training sample will be essentially identical to a non-accident, gaseous sample having a smaller total volume and isotopic inventory than most samples acquired and analyzed on a routine basis by SQN Radiochemistry Laboratory RCL) personnel.
To ensure 10CFR20 limits for dose to personnel are not exceeded, standard radiological protection practices shall be followed (RADCON support when
- needed, use of RWP,. personal dosimetry, etc.).
AAOOOS0 G 0005 CG 000054
)CNI
~-~96 Page 33d i
_of SAFETY EVALUATION 30 DC1 H-06299B ii)
In an accident case, a broken vial releases 15 cc of undiluted RCS off-gas to the PSF atmosphere where it is diluted and vented to the PASS 1iVAC.
If dropped in the CL, it will be diluted by the operation of the exhaust hoods.
SCBA worn by the operators provides protection from ingestion of the airborne contamination.
Operators can vacate the area until the airborne contamination levels lave been reduced by the 11VAC and the dose rates have been assessed by RADCON.
l For training cases, a broken vial presents an airborne release in the PASF sampling operator area.
The PASF 11VAC will be in service and the contents of the 15 cc vial will be diluted and vented from the sampling operator area.
To ensure 10CFR20 limits for dose to personal are not
- exceeded, standard radiological protection practices shall be followed (RADCON support when needed, use of RWP, personal dosimetry, etc.).
Sampling operators shall immediately leave the area and request RADCON assistance.
B.7 The proposed activity does not reduce the margin of safety as defined in the basis for any Technical Specification. No T/S impact was determined to exist.
T/S 6.8.5.e and Licensing Condition 2.F address Post Accident Sampling.
T/S 6.8.5.e requires the plant to have te capability to obtain and analyze various samples under accident conditions.
This modification does not affect the PASS's ability to perform its original design function: diluted RCS off-gas samples will still be capable of being taken.
The modification will allow undiluted RCS off-gas samples to also be taken.
The Licensing Condition 2.F addresses the requirement for having a Post Accident Sampling System, but this modification does not affect, remove, or change the PASS except as stated above.
Nothing in this modification affects either TS 6.8.5.e or Licensing Condition 2.F.
K>
AA000162 CG 000055
vCNI M-0 zq, Page 23
_ oi A~T SAFETY EVALUATION DCN
-06299B
- T/S 3.3.3.7 for Post Accident Monitoring (PAM) requires Type
'j A and/or Category 1 PAM variables to be addressed. Since the
.1 PASS is Category 3, this T/S is not affected. (Ref A.2.v)
T/S indirectly related to this modification are:
1.
2.
3.
4.
This item deleted by Revision 1.
3.6.1.1 - Primary Containment Integrity.
3.7.7 -
Control Room Emergency Ventilation.
3.7.8 -
ux Bldg Gas Treatment System.
tione of the Bases of the above ndirectly related T/S items are affected by the modification.
44G00163 CG 000056 K>_
ADD NEW C
'iECTION Ott PASS PANEL TO ALLOW AN UND_-*UTED SAMPLE OF REACTOR COOLANT OFF-GAS TO BE TAKEN DCNI i SPECIAL REQUIREMENTS R UVERIFIED ASSUMPTIONS U
~~~~~~~~~~Page 31
- 1.
VERIFICATION OF TE TUBING, VALVE, AND SUPPORT INSTALLATION REQUIRES NE-CEB APPROVAL PRIOR TO DECLARING THE SYSTEM OPERABLE.
TO ASSIST IN INSTALLATION, ATTACHMENT A DEPICTS AN ISOMETRIC OF TE MODIFICATION AND THE PLUG VALVE, AD THE PARTS FOR TE VALVE.
NE-CEB VERIFIED TUBING, VALVE /I AND SUPPORT INSTALLATION:
A N E-CB REP.
DATE
- 2.
TOTAL MISSION DOSE CALCULATION (TI-RPS-146)
IS BEING REVISED DUE TO CAQR SQP 90 0191 BEING WRITTEN AGAINST IT. THIS MODIFICATION HAS AN UNVERIFIED ASSUMPTION (UVA)
AGAINST IT IN THAT THE TOTAL DOSE FOR SAMPLE ACQUISITION WILL NOT IMPACT TOTAL MISSION DOSE IN SUCH A MANNER THAT THE SR TOTAL DOSE LIMIT IS EXCEEDED AS CALCULATED BY TI-RPS-146. UNTIL THE UVA IS RESOLVED, A HOLD ORDER IS TO BE PLACED O INSTALLED VALVE RC-V-23.
- 3.
THIS. STEP DELETED PER DCNI F-07036.
- 4.
A HOLD ORDER IS TO BE PLACED ON VALVE RC-V-23 UNTIL TRANSPORT CONTAINER AND HANDLING TECHNIQUE ISSUES ARE RESOLVED.
HOLD ORDER CANNOT BE RELEASED UNTIL SPECIAL REQUIREMENTS OR UNVERIFIED ASSUMPTIONS IN STEP 6 IS COMPLETED.
via HOLD ORDER PLACED ON VALVE# RC-V-2 l
COG. EGR.
DATE
- 5.
NORMAL SAMPLING IS NOT ADDRESSED IN TIS DCN OR IN THE ACCOMPANYING SA/SE.
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SIGNIFICANT CORRECTIVE ACTION REPORT SCAR o. S S C A 9 2 0 0 0 4 Rev. No. 0
~~~~~~~~ ~~~Confliiuiti SheU -
5F E X7 l 5' I 'f"I CONTINUED Step No.
Ib 411 NOTE: Entries made on this sheet shall be signed and dated.
In reference to tem 7 of Corrective Action Plan SQSCA920004
- 7. Perform a study of dilutor valve performance to establish actual dilution factor.
To be cpleted by 10/1/92 -
OT-CEM.
Study o U perFormed on 910/92. Results acceptable.
Study of U2 erfonmed on 9/22/92.
Results unacceptable.
U2 PASS ventilation R COB25Z2 ipeded sanpling. instrument (ion chromatograph) problems suspected as source of U2 analysis data reduction being unacceptable. Request closing of Item for U and extension for U until R worked and re-sanpnlng and re-analysis can be perfonmed.
Extension date of
//,1l71/92 requested.
Signature/Date Title p-v
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L020201/996/9 AA000167 CG 000060
- 1. I-Z SIGNIFICANT CoiuRECTIVE ACTION REPORT SCAR O 1 4
- 2.
0 Rev. [T1 Identify the information that is being continued on this sheet (or example: Decription of Condition).
NOTE: Entries mnde on this shcet sall be signed nd dted.
t52( {E2U~t~Fey=Cl-57
( (5rt Chemistry retuests tint the lROI de d:ite fr develoin a correclivc action pln be extended frnul 6110192 In 6U192. lltis exeltiill i tcuestioe hecu.s oige rlipsmtrv in;tltv1tual nuairpl lo perrnrnm rott cntime nn:lym hi; tntly een nvilile t nddres tis SCAR since 6/5/92. Tis individual urnst I3nve sufliciet lime to j,'in knovledge of lhe PASS nnd related requirements.
perrorin thurougl ivesligntin/roui cse nitilysis, and rep:re n report orrall fndings.
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SIGNIFICANT CORRECTIVE ASTION REPORT SCAkNo.S I10S IC k 1 lOO I l Rev. ~1 Contifilon Shet llrll -
7 t1,117 CONtIUED NOTE: Entrie made on this sheet shall be signed and dated.
Step No.
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SGNIFICAN1 CORRECTIVE ACTION REPORT SCAR No. S S C A 9 2 0 0 4 Rev.
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/ 7 NOTE: Entries made on this sheet shall be signed and dated.
U2 PASS:
ension of item 7 diluter valve study is required due to ventilation system inaperability.
.C075095 and C 03668 have been written. Study for U will be erformed when Whs have been
-ked and system has been restored to operatlonal status. Extension requested for due date I1
/93.
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'LZ0ZOI/996/10 CG 000063 AAOOO7O
INVESTIGATION REPORT FAILURE TO MEET NUREG 0737 REQUIREMENTS FOR POST-ACCIDENT SAMPLING Pnge I or 3 1.0 Problem Description 1.1 Boron analyses/netliod cannot be Performed in three hours or ICas even with the most proficient personnel. NUREG 0737 requires Post-Accident Sampling System (PASS) sample analysis within 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> after a decision is made to collect samples. Approximately 25 %o of personnel are capable of meeting time restrictions of three hours or less excluding boron analysis.
1.2 Number or personnel cnpalie or perroriming PASS sampling/analyses is not sufficient to meet freriuency and duration required by NUREG 0737 (I analysislday for 7 days ollowing accident).
1.3 Diluted PASS samples do not meet comparison crileria when compared with hot sample room results. NUREG 0737 requires that PASS reactor coolant aind containment atmosphere samples he representative of the reactor coolant in the core area and containment atmospIere following an accident or transient.
2.0 Scope or Investigtion A rnot cause analysis or the problems staled above was performed using liazard-Barrier-Target analysis and Causal Factors charling. Inrorination was obtained through interviews with persnnnel experienced vitlh PASS design and operation. dose calculation documentation, and review or requirements.
3.0 Results of Investigation The fllowing seclions contain pertinent acts regarding the three problems staled in Section 1.0.
3.1 PASS Boron Analysis PASS boron analysis is rerrormed using ion chromatography.
The inn chromatongrapli utilized to perrnrm PASS boron analysis is normally configured to analyze secondary systc:in cations.
Reagents must he prepared prior to instrument operation (30 o 45 minutes).
Minor replumbing of the instrumtient is required to switch from cation analysis to PASS boron analysis (20 In 30 minutes).
Analysis of one sample requires a 15-minute nn.
Instniment tist stabilize or at least 30 minutes prior to an initinl run.
3 standrds mist le analyzed to develop n calibratinn curve. and I check atandard mutit ho analyzed prior (n PASS sample analysis (60 tn 75 minutles).
A dedicated instrminent or alternate methoid for PASS analysis is not currently available.
CG 000064 AAOOO17-
INVESTIGATION REPORT FAILURE TO MEET NUREG 0737 REQUIREMENTS FOR POST-ACCIDENT SAMPLING Page 2 or 3 3.0 Restills or Investigation (continued) 3.2 Personnel Technical Specification recdion 6.8.5.e requires that cpahility o obtain and analyze PASS reactor coolant he ensured trnugi training, procedures, and maintenance programs.
25%' or laboratory personnel are proficient in PASS samplinglanalysis.
PASS training and exercises have been repeatedly cancelled or delayed.
Two to rour Chemistry personnel re required to support PASS sampling/analysis depending on the mission.
PASS mission dnse cakulations (1-RPS-146/NEA 3026) indicate that two to three personnel would receive near maximtum allowable dose rar each PASS mission.
PASS mission dose calculations indicate that the majority or dose is oblained during travel between he PASS facility. radiocheinical laboratory, and Operation Support Center (OSC).
Studies have shown that initial dose rates should decrease significantly over time.
3.3 Accuracy or PASS sample Analysis Total gamma activity TGA) has been historically uased to verify PASS sample accuracy.
TGA has a wide agreement range (0.5 to 2.0) when compared tn Hot Sample Room analytical results.
Other PASS sample analytical parameters have a much tighter agreement range than TGA.
PASS dilutor valve dilution ractnr was assuned to he exactly 1:1000.
Actual PASS dilulor valve dilution actor has never been verified.
4.0 Conclusions Based an the evidence presented in Section 3.0. the ollowing conclusions may be drawn:
4.1 PASS Boron Arnalysis The instrument is not dedicated to PASS boron analysis, resulting in longer set-up and analysis time.'
Instrument availability and annlysis time wcre inadequniely assessed in regard to the 3-hour time limit.
CG 000065 AA000172
INVESTIGATION REPORT K>,
FAILURE TO MEET NURECI 0737 REQUIREMENTS FOR POST-ACCIDENT SAMPLING Page 3 of 3 4.0 Conclusions (continued) 4.2 Personnel The PASS raininglexecercise program does not ensure personnel proficiency.
Personnel prnficicncy wa arected by repealed cancellation or delay of activities designed to increase and mainlain rroficiency.
Plant design contribte.s to higl pro ected dose, resulting in quick hum-out' of qualified personnel.
Dose rate changes over time were nnt previously considered in determining number of personnel required to meet the NUREG 0737 sampling requirements.
4.3.
Accuracy of PASS Sample Analysis k
The parameter used to verify PASS sample accuracy has inadequate precision.
The actunl PASS smple dilution factor could not be properly verified due to the inaccuracy or the parameter chosen for verification.
5.0 Causes 5.1 Basic Causes Inadequate niethodlequipient for PASS 1horon analysis.
Inadequate raining/practice.
Inadequate long-term dose sses.sment.
Low management priority.
Inndequate verification nr PASS sample dilution.
5.2 Root Cause Inndequale studylreview.
CG 000066 AA000173
PASS boron cannot he analyzed within 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />.
Improper method Analysis time too long nadequate time study Personnel not proficient Inadequate training Inadequate qualification Insiifficient practice itindequtte prngrnni to m niitnin proficiency Inadequate requirements Low prinrily Inadequate instruction Inadequate instructor Inadequate lesson plan Preparer not qualified Inadequate equipment Equipment not dedicated Insufficient equipment Insufficient budget bindequnte equipment sludy Barrier Method Personnel Equipment Analysis time Time study Qualified evaluator Tltorouglhness Training Practice Qualirication program Instructinn skill Lesson plan Program Priorities Requirements Requirements source Management practice Instnictor qualification Instructor evaluation Preparer qualification Thnroughness Preparer evaluation Dedicated equipment Sufficient equipment Priorities Budget Equipment study Manngement practice Evatiutor qualification Thrnrnughness Jarrel NUREG 0737 requirement Method Analysis time Time study Personnel Training Qualification program Practice Program Requirements Priorities Instruction skill Instructor qualification Lesson plan Preparer qualification Equipment Dedicated equipment Suficient equipment Budget Equipment study AA0001 7 4 CG 000067
Hazard insurficient petsnnnel tn meet NUREG 0737 sampling requirements Insurficient qualified persminct DOse ltimit reached No. personnel inadequate Insuricient manrnwer Si Inacleqtuate facility design Inadequate design study LUniled npliasis PASS not original plant design PASS is etrorit design (not optimum)
Dose rate decrease not considered rver duration of PASS sampling.
Iindrlcunle Insig trni dose rtle sludy Barrier Qualified personnel Training Dose ate No. personnel involved Facility design Dnse rtes decrease Manpower Mnnagcnient policy lidget Design sdy Options Designer qualilications Tharoughness Plant design Plant vintage Post -TMI requirements Dose rate study Thorouglness Target NUREG 0737 requirement Qualified personnel Dose rate No. personnel involved Manpower Facility design Design study Options Plant design Plant vintage Dose rate decrease Dose tale study Training (See H-B-T for Boron Analysis)
CG 000068 AA000175
Hazard PASS sample analyses do nnt meet acceptance criteria when compared t RCS sample analyses.
Dilutinn factor not as expected Dilulion ractor nt verified Inaprrptriae acceptance criteria Inappropriale parameter used Insufficient precision Iitzdequale study Barrier Sample collection Analytical methods Accept-nnce criteria PASS sample dilution Verified sample dilution Appropriate parameters Appropriate precision Prper study Tight range or precision Evaluator qmilification Thoroughness Tareet Analytical accuracy Sample collection PASS sample dilution Acceptance criteria Appropriate parameters Appropriate precision Proper study I
CG 000069 AA000176
C C
Sampling method provides \\I incorrectly diluted sample
\\
I I I
METHOD C
kI III II\\
t i
I I
I
\\a, Sampling valve gives
\\I nco c
dil iflo or -
PASS RCS analytical results do not meet acceptance criteria when compared to I
Hot Samole Room RC analytical results I
Dilution factor different /,
than assumed
/
Dilution factor not /
properly verified.
/
Parameter used for /
verification has wide,'
acceptonce ranae
/
/
/
I',
I
/
'I Inadequate evaluation of parameter I
I
/EQUIPMEN
(E (7~~~~~~~~I 2 to personnel receive METHOD high dose when collecting/
analyzing PASS samples
\\
C Method requires 2 to 3 personnel /1 for sampling/anclysis
/
/
Location of PASS/RCL requires several personnel., /
Inadequate design
,/
Retrofit design 1 t lnadequate personnel/
Post-TMI requiremen Dose rates g I
to decrease, Inadequate dose stud I
ssumed nof over ime II I
I I
I.
I I
I I
II I
Insufficient personnel-to meet long-term sampling requirements' per NUREG 0737 1
Loaction of PASS/RCI requires several I
personnel I.
112 to 3 personnel
/
i burned out per mission 0
G)
CC0 0)4o
/7 I
l I
II.
/EQUIPMENt PEOPLE
C-Lenathv stabilization time.
METHOD c
Instrument not set up/
\\,
lnstrument.not dedicated/
lnsufficJnt no. of instrumenr I
Inadequatey study I
I0 JI
'.No reageni prepared Lengthy calibration rocess \\
I I
I I
1 I
I Analysis time too long,'
Inadeauate method, Inadeauate study/
I
.1 I
I I
t InstrVment not /
,-set up
/
I.
i
//EQUIPMENT Inadequate time to dress out,.
samale, and analyze after clock starts, iPASS boron cannot be analyzed within 3-hour time limit.
I No preparation prior to clock start!
No proactive uidelines Inadequate prograr Inadequate study Varying levels of proficiency
/
I I
Inconsistent training, drill schedules/
Low management priority/
/ PEOPLE
C C
0 C
Lengthy stabilization time METHOD Instrument not set u,/
Instrument. not dedicated/
I insufficient no. of nstrumen Inadeqiuatet, study
\\ No reagents preored-Lenathy calibration process' Analysis time oo ona' Too many cal. ooints' Inadequate method/
Inappropriate QC/
Inadequate studyv Inadeauate study' 0
0 0
0
!PASS boron ccnnot be analyzed within 3-hour time limit.
Inadequate time to dress out, I samole, and cnalyze after clock starts II I
i I
II No preparation /
prior to clock start/
No proactive uidelined Inadequate progrard Inadequate study/
Inc t
I III I
I II 06)
C) 0
-4 co
K)
- 6S II.B.3 POSACCIDET SAMPLDlIG CPABILITY Position A design and operational review of the reactor coolant and containment atmosphere sampling line systems shall be performed to determine the capability of personnel to promptly obtain (less than 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />) a samole under accident cnditions without ncurring a radiation eposure to any individual in excess of 3 and 18-3/4 rem to the whole body or extremities, respectively.
Accident conditions should assume a Regulatory Guide 1.3 or 1.4 release of fission products.
If the review indicates that personnel could not promptly and safely obtain the samples, additional design features or shielding should be provided to meet the criteria.
A design and Derational review of the radiological spectrum analysis facilities shall be performed to determine.the caoability to promptly qantify (in less than 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />) crtain radionuclides that are ndicators of the degree of core damage.
Such radionuclides are noble gases (which indicate cladding failure),
iodines and cesiums (which indicate high fuel temoeratures), and nonvolatile isotopes (which ndicate fuel melting).
The initial reactor coolant spectrum should correspond to a Regulatory Guide 1.3 or 1.A release.
The review shculd also consider the effec:s of direct rdiation from piping and components in the auxiliary building and possible ontamination and direct radiation from airborne effluents.
If the review indica.ts that the analyses required cannot K) be performed in a prompt anner with existing equipment. then design modifica-Q.
tions or equicment procurement shall be undertaken.to meet the citeria.
In addition to the radiological analyses, certain chemical analyses are necessary for monitorinn reactor conditions.
Procedures shall be provided to perform boron and chloride cemical analyses assuming a highly radioac::ve initial samole (eculatcry Guide 1.3 or 1.4 soure trm). Both analyses shall be caoable of being ccmoleed rcmptly (i.e.,
he boron samole analysis within an hour and the cloride sample analysis within a shift).
Chances to Previous Reouirements nd Guidanca This ruirement as originally issue. to all coertting plants by lett2rs dated Seot2mber 1 and Ocober 30, 1979.
Significant changes In rquirements or guidance are:
(1) A;lcws comoir.ed time of 3 houri or le;s for samoling wil analysis.
(21 Soeti!les t-at lc2nsee may usa cnlne sanoling and analvsis t meet h:.e 3-hour time reouirement but mus:
r-v 4de cability t
rve ri soles cf rac:cr :oi2nr and containment a:osnner? for sar2t analysis.
(7) tmo!ement-2tion date has been chanced z january 1 1.
(') Provicer.tesic. cuidance or sr.clno ana analytical caoilit'.
Clar.`cation The ilicwinc ins are :larifica:i.ns f rJirmen:s dentified in lUR~E.-078.
NlUREG-5660, or the Seotemoer 13 and Oc:tcer 30.
79 clarifiction lettari.
i+/-.a._.-5 CG 000074 AA000181
(1) The licensee shall have the capability to promptly obtain reactor coolant samples and containment atmosphere samples.
The combined time allotted for sampling and analysis should be hours or less from the time a decision is made to take a sample.
(2) The licensee shall establish an onsite radiological and chemical analysis capability to provide, within the 3-hour time frame established above,
- ~ quantification of the following:
(a) certain radionuclides in the reactor coolant and containment atmosphere that may be indicators of the degree of core damage (e.g., noble gases; iodines and esiums, and nonvolatile isotopes);
(b) hydrooen levels in the containment atmosphere; (c) dissolved gases (e.g., Us), chloride (time allotted for analysis subject to discussion below), and boron concentration of liquids.
(d) Alternatively, have inline monitoring capabilities to perform all or part of the above analyses.
(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 (RWCUS)] to be placed in operation in order to use the sampling system.
(4)
Pressurized reactor coolant samples are not required if the licensee can ouantify the amount of dissolved gases with unpressurized reactor coolant
.samples.
The measurement of either total dissolved gases or H gas in reactor coolant samples is considered adequate.
Measuring the °2 concentra-tion is recommended, but is not mandatory.
(5) 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.
()
The design basis for plant euipment for reactor coolant and containment atmosohere samnling and analysis must assume that it is possible to ebtain and analyze a sample without radiation exposures to any individual exceeding the criteria of GC 19 (Appendix A, 10 CFR Part 50) (i.e., 5 rem whole body. 75 rem extremities).
(Note that the design and operational review criterion was changed rom the operational limits of 10 CFR Part 20 (AUREG-0578) to the GC 9 criterion (October 30, 1c79 letter from H. R.
Ddnton to all licensees).)
(7) The analysis of primary coolant samples for boron is required for PWRs.
(ate that Revision 2 of Regulatory Guide 1.97, when issued, will likely specify Age need for primary coolant boron analysis capability at WR plants.)
II..3.
1 Z3-67 CG 000075 AAQ008Z?
(9) If inline monitoring is used for any sampling and analytical camability soecified herein, the licensee shall provide backup sampling through grab samples, and shall demonstrate the capability of analyzing the samples.
Established planning for analysis at offsite facilities is acceptable.
Equipment provided for backup sampling shal.l.be capable of providing at
'fit~ t-~gt~*it~
~
l~ I f~tfi5: ~1uir at",
(9) The licensee's radiological and chemical sample analysis capability shall Include provisions to:
(a) Identify and quantify the isotooes of the nuclide categories discussed above to levels corresponding to the source terms given in Regulatcry Guide 1.3 or 1.4 and 1.7.
Where necessary and practicable, the ability to dilute samples to provide capability for measurement and reduction of personnel exoosure should be provided.
Sensitivity of onsite liquid samole analysis cability should be such as to permit measurement cf nuclide concentration in the range frem approximately 1 Ci/g to 10 Cf/g.
(b) Restrict background leveis of radiation in the radiological and chemical analysis facility from sources such hat the sample analysis will provioe results with an accivtably sall error (approximately a factor of 2).
This can be ac:cmolished through the use of sufficient shielding around samoles and outside sources, and by the use f ventilation system design which will control the presence of airb-orne radioactivity.
'!0) Accuracy, range, and sensitivity shall be adecuate to provide pertinent data to the operator in order to describe radiological ano cemic2l status of the reactor coolant systems.
(11) In the design of the postaccident samoling and analysis coability, consideration should e given to the following items:
ta)
Provisions for purging samo:e lines, 'or rducino plateu. in sample
- lines, or minimizing samole los; or distortion, for preventing blockage of sample lines by loose material iii the RCa or cntainment,
'or aorccriate disposal of tie samoles, an-r lcC restrictions to limit ractor _zolant loss frcr a ruotur! of t.: ;Amcle line.
The ost3crident reactor c:olant and containment atmosphere samples should be rresenttve of the reactor coolant in the core rea and the containment atmosohere fol'-c;ino a transien: r accicent.
T'o samole lines should e as short s possible to minimi:e the volume if fluid to e taken from ontainment.
7he residues of sole-collection should e returned t: =:ntainment or to 4 closed system.
(b) T4 ventilation exhaust from he sarnoling staticn should e filtered
- with carcoal adsorners and hn-efficiency artfculata air (f.'n4) ri ters.
(c) Guidelines -3r analytical or instrumentazio. rance are,iven elow in Tble i..3-1.
!!.'.7.7-i
¶CG 000076 AA000183
[.-l-
=216:02 D1 IU WESTERI p.J0.
ItWT.
IU
'93437. '15 P. C2 THREE HOUR OST CCIDEr SA?.Ll;T. AD ANTLYSIS REQUIREMENT The cpability for AS sampling and analysis should be 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> from the time a decision is mde to take a sample.
The sample doesn't have to be taken imediately following an accident.
The 3 hour3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> criterion can start at the time when the sample is initially collected.
It is acceptable to use several hours completing laboratory, countroom and PASS preparations, including nstrumentation checks and sample recirculation, prior to sample collection.
The 3 hour3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> criterion applies to taking one sample ad completing all of the following analyses: radionuclides and hydrogen (if containment atmosphere is sampled) or radionuclides, dissolved gases (e.g.
- 2) nd boron (if reactor coolant s sampled). The 3 our criterion doesn't imply that the PASS must be available iOO% of the time during normal operations.
The PASS is not a required safety-related ystem.
However, PASS systems should be functional for performing its ntended function for any type of accident.
Meeting PASS Program Technical Specifications (Section 6.8.5.e) will ensure that personnel are PAS trained, PAS sampling and analytical procedures are established ad PAS equipment, facilities and supplies are maintained.
- FUREG-0737, Page II.E.3.1-1 provides a requirement change that "Allows combined time of 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> or less for (PAS) samDling and analysis."
NUREG-0737, Page I.B.3.1-2 states under Criterion (1) that The licensee shall have the capability to promptly obtain reactor coolant samples and containment atmospere samples.
The ombined ime allotgf for smrlly arxnalsls h0Uld be 3hours or les Irmfnethe-al~iti~a~ts made to tAke a smp1.
I ITUREG-0737, Page I.B.3.1-2 states under Criterion (2) that *'The
'slicensee shall establish an onsite radiological and chemical analysis capability to provide, Ylthin the 3-hour time frame established aboye, quantification of the following:
(a) certain radionuclides in the reactor coolant and containment atmosphere that may be indicators of the degree of core damage...:
(b) hydrogen levels in the containment atmosphere:
(c) dissolved gases (e.g.,
2), chloride (time alloted for analysis subject to discussion below), and boron concentration of lquids."
r.
REGULATORY GUIDE 1.97;,Rev. 2 states n footnote 17 that IXhe timefor t^.ing andjtoplvzin-samples hold be 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> ** le.s zrpm th:_ii is to samol, except for chloride which should be within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />."
CG 000077 AA000184
.*I-
- 1=
i - --
i -7 : v_-
.11 I ' I
- L.
I Page 2 SQR TECHNICAL SPECIFICTIOTS, Section 6.8.5.e states that "A program which wvii ensur. the capability to obtain and analyze reactor coolant, radioactive lodines and particulates in plant gaseous effluents, nd containment atmosphere samples under accident conditions. The program shall include the following:
(I)
Training of personnel.
(il) Procedures for ampling and analysis, (Lii)
Provisions for maintenance of sampling and analysis euipment."
- 3. CL1RITICATOT OF THE 1UREQ-0737 3 OUR SAMPLING W$D AALYSIS REIJEW_
PROVTIED BY MR KA1s WITT (RC RR) I T
OCTOBER 3
- 4. 19UtS SENTRY PAS OFUR'S GoSIP EETING qT0Es Page 4S -
The "3 hour3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> time limit...for sampling, transport to the lab for analysis s for one sample.
ither RCS sample or containment atmosphere sample.
It's not for both....When does the clock setart?....This sample will not be taken right after the accident happens.
It may be a few ours afterward. I would expect that the technician would be getting ready to take a sample. Getting...suited up and maybe getting circulation for the system established. I would not nclude this in the 3hour period. The period starts when they have actually started taking a sample."
Page 47 -
" wouldn't (consider sample purge time as part of the 3 hour3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> criteria).
I think that's in preparation to take the sample."
- Attachment after page 6l,"Counterpart Meeting Summary On PASS...April 24-26. 1984", page 14 -
"The PASS is operable when all samples and mesurements specified n IIB.3 can be bthined within the 3 hour3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> time f £rame."
Attachment fter page 61,"Counterpart Meeting Summary On PASS...April 24-26. 1984", page 15 -
uegt:qR Does the requirement to obtain and analyze a sample ithin 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> imply 100% availability of the PASS during normal operations and any type of potential accident? Answer 11o, however, PASS systems should be functional for performing intended.
functions for any type of accident, when required.
Attachment after page 61,"Counterpart Meeting Summary On PASS...April 24-25.
1994", page 18 - "XQi~tjt Does the 3 hour3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> time constraint for sample acquisition and analysis apply to both containment air and reactor coolant or just reactor coolant?
Answer! The core damage estimate procedure specifies the prime sample location which takes nto account the type of accident and here the most representative sample of core conditions should be obtained.
Therefore, the 3 hour3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> time constraint for sample acquisition and analysis is for the one sample (either reactor coolant or containment atmosphere) which is most representative of core conditions.
The econd sample must also be taken and analyztd, but not within the 3 hour3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> time limit."
o0<<>S~t t
AA00018s CG 000078 g,
Don E. Adams I U*_ a 1
al
U13 920707 801 July 7, 1992 J. M. Stitt, Sequoyah CA Coordinator, SB1C, SQN BELLEFONTE NUCLEAR PAST (BLN),
SITE QUALITY, GENERIC APPLICABILITY REVIEW OF CORRECTIVE ACTION DOCUMENT.
SQSCA920004 has been reviewed for applicability to BLN.
It has been placed in inactive status and added to generic collector BLPER910008. Thank you for the information. Please update TROI to reflect closure of this action.
If you have any questions call me at 574-8926.
Dennis Sanders Correhtive Action Coordinator OSE-l',
LN cc:
TV*A 64 tO'-!-65) (op-wr-5.s5 t'NrrI.n STATES G;OlMINMENT Kj1 el1) or lll tum TEDNNESSEE VALLEY AUTHORITY T42 9208 07 9 5 6
- R. M. Norton,' CAQ Coordinator, NA, Watts Bar Nuclear Plant, DATE
- 7
SUBJECT:
WATTS BAR NUCLEAR PLANT (WBN) -
ADVERSE CONDITION GENERIC REVIEW RESULTS NUMBER
`2)5C59ci°°O 4 Attached is the result(s) of WBN review for generic applicability to this site.
Please have IROI updated to indicate action closure.
- !a I
t o nL-R.. Norton DAK Attachment
(
AA0OOO187C 008 CG 000080 I14/1
Subject SCAR/FIR:
____cA9_Zc_
Assigned to 01 Cv D
O VA
/ 7-/0 Date Due Date
_Z rlease review the subject adverse condition for generic applicability to Watts Bar.
Confirmed Applicable Yes -
l b" ( X )
Other
)
If no or other, provide justification:
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yes, a tech a copy of the resulting or previous adverse cndition.
A Aver -ondition numbers:
Prepared By:
L U
Approved B:
. \\2v
/
7 /3t ate
/.-,/ / t Dnat e AAO00188 CG 000081 2M/47
SIGNIFICANT CORRECTIVE ACTION REPORT SCARNo.
0 C-12°OO6 Rev.
J a.
R.w 7t Q
ldcnliry the inrormation that is being continued on this sheet (for example: Description o[ Condition).
NOTE: Entries maldc on this shect shall be signed and dated.
BROWNS FRRY'S GENERIc REVIEW OF SQSCA920004 Rev.
o Response to Requirement Violated and Description of Condition:
Item t,1: Browns Ferry Chemistry personnel can take and analyze any samplg. within three hours from the time a decision is made to sample.
This is demonstrated during REP drills with the most recent being June 3, 1992.
Boron analysis requires a setup time of approximately 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and an analysis time of approximately 15 minutes.
Since the boron analysis setup time is in parallel with the sampling time (approximately 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />), the sampling and analysis can be completed within 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />.
Item 2:
Browns Ferry's Post-Accident Sampling System (PASS) is a grab sample panel.
Time-motion studies show the dose to sample and analyze all points is-approximately 1.3 Rem whole body and 11.4 Rem extremities.
The maximum allowable dose to any individual for PASS is 5 Rem whole body and 75 Rem extremities.
With the current fourteen trained RLAs, Browns Ferry can take and analyze one sample per day for seven days and one sample per week for at least 30 weeks without exceeding the dose limits. This is more than adequate time to ensure accident conditions no longer exist.
Item 13:
Browns Ferry uses 0-TI-222 on a quarterly basis to compare PASS samples to normal samples.
Comparison of samples to date have met acceptance criteria.
KeIth smith Ext.
7/zz.q
- 1. Da e Ai A 000189 CG 000082