ML20129H882
Text
- - -
'l For:
The Commissioners EtQm:
' James M. Taylor Executive Director.for Operations Sub.iect:-
PROPOSED DIRECTOR'S DECISION REGARDING A MATERIAL FALSE STATEMENT DURING A COMMISSION j
MEETING ON V0GTLE UNIT 2
Purpose:
To consult with.the. Commission concerning whether or not to take enforcement action for inaccurate information provided by an official of the Georgia Power Company (GPC) during a March 30, 1989, Commission meeting to consider issuing the full power operating license for Vogtle Electric Generating Plant (VEGP) Unit 2.
.i Discussion:
Ori March 30, 1989, the Commissioners met to discuss and possibly vote on the full power operating license for VEGP-Unit 2.
The Commissioners present were' Chairman Lando W.
Zech, Jr., and Kenneth M. Carr, Thomas M.
Roberts, Kenneth C. Rogers, and James R.
Curtiss.
The transcript (Enclosure 1) reflects that then Commissioner Carr expressed concern about the hierarchy between the Vogtle plant manager (i.e., the General Manager) and the Chief Executive Officer (CE0), noting that it l
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Contact:
David Matthews, NRR 492-1490 gIY j6] 10 g 1 960827 5, KOHN95-211 PDR
O The Commissioners,
" looked to me like he was a long way from the CEO."
Mr. R. P. Mcdonald, GPC Executive Vice President - Nuclear Operations, responded that (1) he (Mr. Mcdonald) reported to Mr. Bill Dahlberg, the GPC CEO, (2) that Mr. Ken McCoy, Vice President of Vogtle, reported to Mr.
Mcdonald, and (3) that Mr. George Bockhold, then Vogtle General Manager, reported directly to Mr.
McCoy.
At the conclusion of the meeting,the Commissioners voted unanimously in favor of the license, and the. license was issued the following day.
On May 1,1989, Mr. W. G. Hairston, III, Senior Vice President for Nuclear Operations, forwarded to the NRC a letter of correction of the i
transcript (Enclosure 2), noting that Mr.
McDor. eld had inadvertently left out the Senior Vice President of Nuclear Operations.
The organization is as described on figures 13.1.1-1 and 13.1.1-2 of the Vogtle Final Safety Analysis Report.
On September 11, 1990, a Petition to initiate proceedings and impose civil penalties was filed by Michael D. Kohn, Esquire, on behalf of Petitioners Marvin B. Hobby and Allen L.
Mosbaugh (Enclosure 3).
The Petition claimed that Mr. Mcdonald knowingly made false statements to the NRC Commissioners in the presence of Messrs. Dahlberg, McCoy, and Bockhold during his response to then Commissioner Carr in that he " eliminated one entire level of management between the plant manager and the CEO."
Moreover, the Petition asserts that Messrs. Dahlberg, McCoy and Bockhold should have known that Mr. Mcdonald's statements were false and should have brought
Contact:
David-Matthews, NRR 492-1490
t i
The Commissioners :
this to the immediate atte '*on of the Commission and otherwise corrected the record before the Commission acted on the Vogtle full power license request.
By letter of September-21, 1990,. Mr.- Kohn- -
forwarded exhibits to support the Petition, including relevant pages of the transcript of the March 30, 1989, meeting.
The issue was j
renewed in a Supplement to the Petition of July 8, 1991 (Enclosure 4).
In the Supplement, the f
Petitioners noted that, even though'Mr.
L Mcdonald's reply contained an omission, it i
stil1did not satisfy the concern of Commissioner 1
Carr who subsequently replied, "I still have my concern, I guess."
l GPC responded to the Petition on April 1,1991 1
(Enclosure 5).
The response noted that the Commission had been. apprised of the Company':
organization before the proceeding on March 30, including the Senior Vice President position, by o
i an amendment to the VEGP Final Safety Analysis Report (FSAR) that was submitted November 23, l-1988.
The amendment described the reporting chain from Mr. McCoy to Mr. Wairston to Mr.
Mcdonald.
GPC's response also stated that the j.
NRC had reviewed the organizational structure in F
December 1988, and issued an inspection report.
The inspection report stated that the vice presidents of the Farley, Hatch and Vogtle projects reported to the Senior Vice President who reported to the Executive Vice President, and that the organization was consistent with the-VEGP-FSAR amendment submitted in November 1988.
i
.The reply by GPC also noted that, during the March 30, 1989, proceeding, Commissioner Rogers stated that he had reviewed the Company's organizational chart during a visit he made to the plant site.
P
O.
The Commissioners 3 The written response by GPC to the Petition also
' notes that GPC had submitted the letter of correction of the transcript approximately 2 weeks after receiving the NRC transcript.
The NRC staff has reviewed this issue and has 3
concluded that Mr. Mcdonald's reply to then Commissioner Carr was inaccurate in that the transcribed record clearly contradicted other documents of record, including the FSAR and NRC i
inspection reports.
The failure was material in i
that the reply was in direct response to the Commissioner's stated concern for an organizational structure in which the plant manager appeared to be "a long way from the CEO," and could influence the Commission, and may have been considered by the Commission in j
its decision.
The staff should have recognized the inaccuracy of the statement made by Mr. Mcdonald in that it had reviewed the licensee's submittal.
The staff did not correct the misstatement at the time it was made.
It is not clear if the staff heard the specific exchange or, if it did, whether or not the staff thought the statement was inconsequential.
The licensee or its employees would not likely attempt to deliberately mislead the Commissioners since the licensee had previously provided correct information and NRC staff members were present who knew the correct information.
Therefore, the staff believes that Mr.
Mcdonald's failure was not intentional.
One issue is what action should be taken for an i
unintentional oral statement made by a licensee official that was not corrected before it was relied upon.
Under the Enforcement Policy, normally unsworn oral statements that are unintentionally inaccurate are not acted upon
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The Commissioners,
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unless they involve significant information by a licensee official.
The Policy recognizes the i
reliability issue-associated with oral i
statements, but states "The Commission must be
.able to rely on oral communications from licensee officials concerning significant information"-[10 CFR Part 2, Appendix C, Section VI].
Therefore, under the Enforcement Policy, the NRC could issue a severity level III violation for Mr. Mcdonald's statement if it were considered significant.
If it were not significant, then no enforcement action would be appropriate.
The second issue is what action should be taken with regard to the failure of the other GPC personnel present during the Commission meeting to correct Mr. Mcdonald's statement at the time it was made.
Given that the Commission was to vote that day whether or not to authorize issuance of an operating license, it can be argued that GPC personnel should have corrected the false statement immediately if they were aware of it.
Tha staff does not know whether or not GPC personnel recognized and knowingly allowed an omission in this matter.
The staff believes that, while the statement (and thus the omission) were material since they could have influenced the Commission, they were not significant because the staff does not believe-this one issue would have caused the Commission to reach a different decision.
However, the Commissioners to whom the statement was made may believe it was significant.
Therefore, unless the Commission informs the staff-that'the matter was significant, the staff will respond to the Petition with the understanding that the matter was not significant.
With regard to the Mcdonald statement, the staff will respond that, although
The Commissioners a false statement was made to the Commission, the staff believes it was unintentional.
- Thus, in accordance with the Enforcement Policy, no enforcement action is warranted due to the lack of significance.
With regard to the material omission, the staff will respond that it does not know if GPC knowingly made an omission in this matter, but that further action to pursue this omission is not warranted due to its lack of significance.
The staff will further indicate in its Director's Decision that c:
Commission has concurred in this approact.
Coordination:
The Office of the General Counsel has no 16, objection to this paper.
The staff has not requested, and the Office of Investigations does not intend to initiate, an investigation into this matter.
=._:.- = a -
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io The Commissioners 3 Recommendation: ' Unless the Commission directs otherwise within 2 weeks from the date of this paper, the staff will respond to the Petition with the understanding that Mr. Mcdonald's failure was not considered significant.
James M. Taylor Executive Director for Operations
Enclosures:
1.
Transcript of Commission Meeting i
1 on March 30, 1989 2.
May 1, 1989, Letter of Correction 4
3.
September 11, 1990, Petition 4.
July 8, 1991, Supplement to Petition 5.
Response by R. P. Mcdonald of April 1,1991, to Petition
e The Commissioners R.ecommendation:
Unless the Commission directs otherwise within 2 weeks from the date of this paper, the staff will respond to the Petition with the understanding that Mr. Mcdonald's failure was not. considered signi.ficant.
James M. Taylor Executive Director for Operations l
Enclosures:
1.
Transcript of Commission Meeting on March 30, 1989 l
2.
May 1, 1989, Letter of Correction 3.
September 11, 1990, Petition 4.
July 8, 1991, Supplement to Petition 5.
Response by R. P. Mcdonald of April 1,1991, to Petition f
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.NOV 13,566 08-07-91 08:10a 920401
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.VOGTLE10.RPT185,217 08-08-91 02:05p VOGTLE20.RPT 124,532 12-16-91 09:37p VOGTLE21.RPT123,952 07-31-91 04:30p VOGTLE8.RPT 190,244 09-28-90 04:53p hNd A
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- MEMORANDUM FOR
- 8. Fiore111. Chief. Reactor Operations and i
Nuclear Support Branch RI!!
FRON:
J. H. Snierek. Assistant Director for Field Coordination. 202/IE j
$U5 JECT:
DPERABILITY DEMONSTRATION OF REDUNDANT SYSTE 1
We have discussed with DDR the issue reised in your memorandum of' 4
April 27.1977. The NRC philosophy of testing redundant systems when i
one system fails is unde eing a change. The current feeling is that i
to take its redundant sys out of service fbr testing if the first l
system fails, creates the risk of the second system also failing. It j
has been obsa'rved that failures of the second system are often esisted i
to the test itself and is not an indication that the system would have j
failed should it have been needed.
f.
All current STS reflect this thinking and some T5 changes am occurring i
to improve older 78. Some older facilities. however, are reluctant to
Initiator 4 was found to be the most limiting event for modes 3, 4, and 5.
The parameters used in the calculation of time l
available for operator response are listed in table 15.4.6-1.
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15.4.6-2a Amend. 17 7/85
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I Amend. 17 7/85 15.4.6-2b l
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VEGP-FSAR-15 O
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15.4.6.2.1.3 Dilution During Full Power Operation, Including Startup.
15.4.6.2.1.3.1 Dilution During Startup.
Conditions at startup E17 require the reactor to have available at least 1.30-percent i
The maximum boron concentration Ak/k shutdown margin.
l required to meet this shutdown margin is conservatively The following conditions are assumed-l estimatied to be 1704 ppa.
for an uncontrolled boron dilution during startup:
Dilution flow is assumed to be the combined capacity A.
of the two primary water makeup pumps'(approximately 242 gal / min).
B.
A minimum water volume (9757 ft*) in the reactor l
coolant system is used.
This volume corresponds to the active volume of the RCS minus the pressurizer l
i volume.
15.4.6.2.1.3.2 Dilution During Power Operation.
During power Il7 operacion, the plant may be operated two ways, under me.nual i
While operator control or under automatic Tavg/ rod control.
the plant is in manual control, the dilution flow is assumed to be a maximum of 242 gal / min, which is the combined capacity of While in automatic the two primary water makeup pumps.
control, the dilution flow is limited by the maximum letdown flow (approximately 125 gal / min).
Conditions at power operation require the reactor to have The available at least 1.30-percent Ak/k shutdown margin.
maximum boron concentration required to meet this shutdown margin is very conservatively estimated to be 1704 ppm.
15.4.6-3 Amend. 17 7/85
VEGP-FSAR-15 0
A minimum water volume (9757 ft*) in the RCS is used.
This volume corresponds to the active volume of the RCS minus the pressurizer volume.
15.4.6.2.2 Results The calculated sequence of events is shown in table 15.4.1-1.
Dilution During Refueling.
Dilution during 15.4.6.2.2.1 refueling cannot occur due to administrative controls.
(See ag paragraph 15.4.6.2.1.1).
11 j
15.4.6.2.2.2 Dilution During Cold Shutdown.
For dilution during cold shutdcwn, the minimum time from the "high flux at shutdown" alarm to the lo'ss of shutdown margin is 15 min.
I Dilution During Hot Standby and Hot Shutdown.
15.4.6.2.2.3 the minimum time For dilution during hot standby and startup, required for the shutdown margin to be lost and the reactor to become critical is 15.3 min.
15.4.6.2.2.4 Dilution During Startup.
In the svent of an g coach to criticality or dilution during power unplanned e-escalation while in the startup mode, the operator is alerted' to an unplanned dilution by a reactor trip at the power range low setpoint.
After reactor trip there is neutron flux high, at least 19.0 min for operator action prior to loss of shutdown
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margin.
15.4.6.2.2.5 Dilution During Power Operation.
During full-power operation with the reactor in manual control, the operator is alerted to an uncontrolled dilution by an At least 19.0 min are c
overtemperature AT reactor trip.
available from the trip for operator action prior to loss of
(
During full-power operation with the reactor in automatic control, the operator is alerted to an uncontrolled reactivity At least 36.8 min C
insertion by the rod insertion limit alarms.
are available for operator action from the low-low rod
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insertion limit alarm until a loss of shutdown margin occurs.
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15.4.6-4 Amend. 17 7/85 h
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VEGP-FSAR-15 o
A 15.4.6.3 conclusions Il7 The results presented above show that adequate time is available for the operator to manually terminate the source of dilution flow.
Following termination of the dilution flow, the operator can initiate reboration to recover the shutdown margin.
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4 0150V Amend. 1 11/83 15.4.6-5 Amend. 17 7/85
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VEGP-FSAR-15 i+
TABLE 15.4.6-1 PARAMETERS l
Dilution Flowrates:
Initiator Flowrate (apm) 1 63 2
120 3.5 3
hS6 4
Volumes:
f Mode Volume (ft3)
Volume (aal) 17 3, 4 5840.0 43800 5affilled) 11200.9 84007 t
1 Sb (drained) 3435.0 25763 Boron Concentrations:
C Co (ppm)
Percent SDM I
Mode e gppm)
J 3, 4 720 870 2.0 5
971 1077 1.5 O
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. Mr. F. 8. Marsh c
5.0. GAE-4704 Project Engineering Manager 2*
Bechtel Power Corporation Ref: VE6P OSER 4toteL@od-Open Item 101 P.O. Box 60860 Terminal Annex Los Angeles, California 90060 V0GTLE ELECTRIC GENERATING PLANT UNITS 1 AND 2 Final Resort for Baron Dilution
Dear Mr. Marsh:
Attached is the final report for the Boron Dilution PRA. Please transmit this to the NRC.
If you have any questions, please do not hesitate to call me.
Very truly yours.
WESTINGHOUSE ELECTRIC CORPORATION I
C.
N J. L. Vota, Manager Southern Company Projects R. J. Morrison/bek/0452n Attachment cc:
F. B. Marsh, IL, 2A
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Vogtle Project July 25, 1985 4
1 Director of Nuclear teactor Regulation Filat X7BC35 Attentions Ms. Elinor G. Adanssa, Chief Lost GN-666 I
Licensing Branch #4 Division of Licensing i
U.S. Nuclear Regulatory Commission Washington, D.C.
20555
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NRC DOCIET NUMERS 50-424 AND 50-425 CONSTRUCTION PERMIT NUMERS CFFR-108 AND CFFR-1209 j
V0GTIE EIECTRIC GENERATING F1 ANT - UNITS 1 AND 2 SER CONFIRMATION ITEM-47: BORON DILUTION i
j Dear Mr. Dentons l
Your staff requested a copy of the Vogtle Electric Generating Plant Boron l
Dilution Analysis. Enclosed are five copies of the requested report. The results of this report have been incorporated in Amendment 17 of the VEGP FSAR.
i If your staff requires say additional information, please do not hesitate to contact me.
i sincerely,
J. A. Bailey l
Project Licensing Manager JAB /cas l
Enclosure ze:
D. O. Foster G. Bockhold, Jr R. A. Thomas T. Johnson (W/o Encl.)
J. E. Joiner, Esquire D. C. Teper (W/o Encl.)
B. W. Churchill, Esquire L. Fowler M. A. Miller W. C. Eassey B. Jones, Esquire (W/o Encl.)
VoEtie Project File L. T. Gucus
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V0GTLE ELECTRIC GENERATING PLANT 80RON DILUTION ANALYSIS I
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TABLE OF CONTENTS
?.AA1 1
1.0 INTR 000CTION.
2 2.0 SYSTEM DESCRIPTION 2
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2.1 Function 3
2.2 General Description 10 2.3 Electric Power Requirements 10 2.4 RMCS Operation 15 2.5 CVCS Operation 21 3.0 ANALYSIS f.
21 3.1 Initiating Events 24 3.2 Operator Response Times 25 3.3 Event Trees 27
[
3.4 Failure Probabilities for Top Events 36 4.0 RESULTS 38 I
5.0 CONCLUSION
S 39 REFERENCES A-1 I
APPENDIX A: Failure Modes and Effects Analysis 8-1 APPEN0!X 8: Response Time Calculations C-1 APPENDIX C: Data i
1 80260:10/061885 if
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V0GTLE ELECTRIC GENERATING PLANT ~
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BORON DILUTION ANALYSIS l
1.0 INTRODUCTION
)
An analysis was used to evaluate burcn dilution events during shutdown modes l
3 through 6 at the Vogtle Electric Generating Plant (VEGP).
The analysis l
j estimated the probab,ility of incurring a boron dilution event which progresses to unplanned criticality. The technique modeled realistic plant conditions l
and msponses, including both mechanical and human errors.
l i
l Failure modes and effects analysis, human error analysis, and event tree analysis were used to determine.possible events and m sponses to those i
Single failure criteria was applied when determining initiating l
events.
Time intervals from alars to loss of shutdown margin were calculated 4 -
events.
l for each initiator in each mode to determine the length of time available for These calculations depended on dilution flowrates, initial l
operator response.
boron concentrations, and Peactor Coolant System (RCS) volumes specific to the l
From these times, human error probabilities were determined.
l
-event and mode.
Mechanical and human failure rates were then assigned to the event trees, and I
the trees were quantified to obtain an estimate of the probability of loss of
{
shutdown margin due to boron dilution. Also, this analysis provided a list of the most likely scenarios leading to this type of accident, referred to as i
j dominant accident sequences.
w The Chemical and Volume Control System (CVCS) is described in the following l
The section to aid in understanding the analysis described in Section 3.
3~
results and conslusions are presented in Sections 4 and 5 respectively.
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-2.-O-CHEMICAL AND VOLUME CONTROL SYSTEM DESCRIPTION 2.1 FUNCTION s
The basic functions of the Chemical and Volene Control System (CVCS) are as j
follows:
d Maintain prograssmed water level in the pressurizer, i.e., maintain 4
a.
requiredwaterinventoryinReactorCoolant;5ystem(RCS).
j b
Maintain seal-water injection flow to the reactor coolant pumps, control reactor coolant water chemistry conditions, activity level, c
soluble chemical neutron absorber concentration and makeup.
1 d.
Provide means for filling, draining, and pressure testing of the RCS.
I Provide injection flow to the RCS following actuation of the Safety e.
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Injection System (SIS).
4 i
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2.2. ENERAL DESCRIPTION Figures 2-1 through 2-4 show the CVCS piping and instrumentation.
l 2.2.1 CHARGING. LETDOWN AND SEAL WATER a
The charging and letdown functions of the system are employed to maintain a programmed water level in the reactor ceplant system pressurizer, thus maintaining proper reactor coolant inventory during all phases of plant ll This is achieved by means of a continuou~s feed and bleed process
?
operation.
during which time the feed rate is automatically controlled h pressurizer The bleed rate can be chosen to suit various plant operational water level.
requirements by selecting the proper combination of letdown orifices in the Reactor coolant is discharged to the CVCS from cold leg of letdown flow path.
the RC5; it then flows through the shell side of the regenerative heat exchanger where, during normal operation, its temperature is reduced to The coolant then experiences a large pressure reduction approximately 2g0'F.
in passing through a letdown orifice (4, = 1700 psi) and after passing through the containment boundary it flows through the tube side of the letdown heat exchanger where its temperature is further reduced to about 115'F.
l Downstream of the letdown heat exchanger a second pressure reduction occurs as I
This pressure the coolant flows to the purification system domineralizers.
reduction is performed by the low-pressure letdown valve, which maintains an upstream pressure sufficient to prevent flashing downstream of the letdown i
orifices.
The coolant then normally flows through one of the mixed-bed domineralizers l
through the reactor coolant filter, and into the volume control tank via a The l
diversion valve and finally a spray nozzle in the gas space of the tank.
j gas space in the volume control tank is filled with hydrogen, which is l
The i
regulated to a pressure of 15-20 psig during normal plant operation.
4 partial pressure of hydrogen in the volume control tank determines the concentration of hydrogen dissolved in the reactor coolant..
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i 80260:10/022285 3
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\\ l l-An alternate letdown path is provided which allows the letdown flow to pass l
through the Boron Thermal Regeneration System (gTRS) when boron concentration j
changes are desired to follow plant load. This alternate letdown flow path is f
After directed to the STRS downstream of the mixed bed dominera111ers.
processing by the BTRS, the flow is returned to the CVC5 at a point' upstream 4
i of the reactor coolant filter.
The charging pumps normally take suction.from the volume control tank and j
return the cooled, purified reactor coolant to the reactor coolant system via j
j the charging system. The charging pumps discharge at a pressure dictated by j
the prevailing reactor coolant system pressure, the resistance of the charging j
line, and the pressure drop impressed by the positioning of an air-operated control valve situated in the charging line (norus11y the pressure will be l
j about 2350 psig). Normal charging flow is handled by the positive j
displacement (PD) charging pump or one of the centrifugal charging pumps.
Should the PD pump be operating the flow rate will be dependent upon the speed I
of the positive displacement charging pump, controlled either by pressuriter l
1evel requirements or by operator choice. If this pump reaches the high speed f
limit, it becomes necessary to place a centrifugal pump in operation to provide the higher flow capacity, and to remove the positive displacement pump l
i from service.
I The flow rate for the centrifugal charging pump is controlled by a modulating a
g valve in the pump discharge line. The charging flow controller maintains the preset charging flow, which is preset by the pressuriter level requirements.
i j
A minimum flow for the centrifugal charging pump protection is continuously t
l diverted from the charging pump discharge via a miniflow orifice and. the seal l
water heat exchanger back to the volume control tank discharge.
t i
The bulk of the charging flow is pumped back to the reactor coolant system via ll.
the tube side of the regenerative heat axchanger where the outlet temperature l
approaches the reactor coolant temperature. The flow is the injected into a cold leg of the RC5. Two redundant charging paths are provided. A flow path is also provided from the regenerative heat exchanger outlet to the h
pressuriter spray line. An air operated valve in the line is employed to l
I F
4 80260:10/022285 e
i s
provide auxiliary spray to the vapor space of the pressurizer during cooldow I
(
to supplement the spray from the reactor coolant system, and thus provide a rapid means of cooling the pressurizer near the end of plant cooldown when th j
reactor coolant pumps are not operating.
4, i
The remainder of the charging flow is directed to the reactor coolant pumps i
f It enters the pumps at a point' between via the seal 9sater-injection filters.
Here the flow splits and a portion s-the labyrinth seal and the No. 1 seal.
l enters the RC5 via the labyrinth seals and thermal-barrier-cooler cavity, with the remainder flowing up the pump shaft and leaving the pump via the No. 2 seal.
I i
The labyrinth flows are removed from the RCS as a portion of the letdown i
f The No. 1 seal discharges flow to a cosmon manifold, exits the i
flow.
containment, and then passes through the seal-1 dater return filter and the seal-esater heat exchanger to the suction side of the charging pumps, or by i
The alternate path enters the alternate path to the volume control tank.
volume control tank through a spray nozzle which is used during plant shutdow and degassing to provide some recycle flow to improve the overall fission gas stripping in the tank.
t An alternate 1.etdown path from the RCS is provided in the event that the Reactor coolant can be discharged from a normal letdown _ path is inoperable.
cold leg and flows through the tube side of the excess letdown heat exchanger,
[
where it is cooled to ~165'F.
Downstream of the heat exchanger a The flow then remots-manual control valve controls the excess letdown flow.
noriaally joins the No.1 seal discharge manifold flow and passes through the seal water return filter and seal water heat exchanger to the suction side f
The excess letdown flow can also be directed to the the charging pumps.
The l
reactor coolant drain tank, bypassing the No. 1 seal return manifold.
excess letdown flow path can also be used to maintain normal heatup rate of the plant, by providing additional letdown capability during the final stages C
This path removes some of the excess reactor coolant due to of heatup.
expansion of the system as a result of the reactor coolant system temperatur increase.
l i
5 l
80260:10/030685
l r
i*
Surges in reactor coolant system volume due to load changes are accommodated f
for the most part in the pressurizer; however, the volume control tank is designed to accommodate programmed pressurizer level mismatches which may occur due to a i 4*F temperature error. High water level in the volume control tank causes letdown flow nornelly entering the tank to be d'Iverted to l-Low level in the recycle holdup tanks located in the Soron Recycle System.
the volume control tank initiates makeup from the Reactor Makeup Control l
1 System (RMCS). If the RMCS does not supply sufficient makeup to keep the l
volume control tank level from continuing to fall, a low-low level signal f
actuates an alars and causes the suction of the char ~ging pumps to be 1
transferred to the refueling water storage tank.
s l
2.2.2 CHEMICAL CONTROL, PURIFICATION AND MAKEUP The water chemistry, chemical shin and makeup requirements of the RCS are such I
)
that the following functions must be provided:
l Means of addition and removal of pH control chemicals for startup and i
a.
j normal operation.
l 4
Control of oxygen concentration during normal and shutdown operation l
b.
l of the plant.
)
Means of purification to remove corrosion and fission products.
c.
Means of addition and removal of soluble chemical neutron absorber d.
(boron) and makeup water at concentrations and rates compatible with all phases of plant operation including emergency situations.
oH Control The chemical control element employed for pH control is lithium hydroxide This chemical is chosen for its compatibility with the materials (Li 0H).
y and water chemistry of borated water / stainless steel / zirconium systems; in addition, Li-7 is produced in the core region due to irradiation of the dissolved boron in the coolant. The Li 0H is introduced to the reactor y
coolant system via the charging flow. A chemical mixing tank is provided to 00260:10/022285 6
i The solution is introduce the solution to the suction of the charging pumps.
Reactor prepared in the laboratory and poured into the chemical mixing tank.
l-f makeup water is then used to flush the solution to the suction manifold of the charging pump.
omveen control
~
j l
During initial plant startup from the col,d condition hydrazine is employed as
'an oxygen scavenging agent. The hydrazine solution is, introduced to the reactor coolant system in the same manner as described above for the pH l
Hydrazine is not employed at any time other than startup from control agent.
l the cold shutdown state. During norinal plant operation, hydrogen in the l
reactor coolant scavenges oxygen produced in the core region due to the Sufficient partial pressure of hydrogen is maintained in i
radiolysis of water.
the volume control tank, such,that an equilibrium concentration of 25 - 35 cc of hydrogen per kg of reactor coolant is maintained in the reactor coolant P
sing y dap u
ontro a e ma a
ressure of 15 to 20 psig in the vapor space of the volume control tank. This regulator i
can be adjusted to provide the correct equilibrium hydrogen concentration.
{
purification Mixed-bed domineralizers are provided in the letdown line to provide cleanup of the letdown flow. The domineralizers remove ionic corrosion products, certain fission products, and act as filters. One dominera11rer is usually in continuous service for nories1 letdown flow and can be supplemented by the The cation cation bed domineralizer when additional purification is desired.
resin removet principally cesium and lithium isotopes from the purification When the BR$ is being utilized, the cation bed domineralizer is flow.
utilized to remove as much Li-7 and Cesium as possible before the water is diverted to the BRS.
b
t i.
3j.
[
n Turther cleanup feature is provided for use during cold shutdown and i
residual heat removal. A remote-operated valve admits a bypass flow from the j
Residual Heat Removal System (RHRS) into the letdown line upstream of the l
1etdown heat exchanger. The flow passes through the heat exchanger.,through a j
mixed-bed domineralizer and a reactor coolant filter to the volume control The fluid is then returned to the reactor coolant system via the normal l
tank.
charging route.
i, Filters are provided at various locations to ensure filtration of particulate I
and resin fines'and to protect the seals on the reactor coolant pumps.
I i
Chemical Shim and Makeun
~
i The function of soluble neutron absorber (boron) concentration control and f
makeup is provided by the RMCS employing 4 wt. percent boric acid solution and In addition, l'
reactor makeup water from the Reactor Makeup Water Storage Tank.
for emergency boration and makeup the capability exists to provide refueling l
l water or 4 wt. percent boric acid to the suction of'the charging pump.
Initial filling and makeup quantities of 4 wt. percent boric acid solution are l'
l.
prepared in the boric acid batching tank where boric acid crystals are The I'
dissolved in hot water and pumped to the boric acid storage tank.
batching tank is steam-heated to allow heating the contents to the desired l
temperature (= 85'F) at which the 4 wt. percent solution is prepared.
i The batch is transferred to the boric acid storage tank by the transfer l
The two tank is located in a compartment that is maintained at a l
pumps.
temperature greater than or equal to 65'F. A pump can be periodically run to 1
recirculate the tank contents through the boric acid filter back to the tank.
l On a demand signal by the RMCS, one pump aligned to that unit where the signal was generated starts and delivers boric acid for makeup.
t The reactor makeup water pumps take suction from the reactor makeup water
{
storage tank, and are employed for various makeup and flushing operations One of those pumps also starts on demand from the throughout the systems.
RMCS.
4 l
80260:10/030885 8
. -.. -. ~. _. -. -...
i The flow of boric acid from the boric acid transfer pump and the reactor askeup water from the reactor makeup water pump is directed to either the l
suction annifold of the charging pumps or is sprayed into the volume control tank through the spray nozzle. The normal flow path will be the line to the volume control tanks where hydrogen pickup will be assured during long In the event that xenon transients require rapit.
j dilution processes.
beration, the direct line to the charging pushs suction can be used.
i l
3 j'
I i
2 i
e ii i
T li
>l 80260:10/022285 9
7 2.3 ELECTRICAL POWER RE00!REMENTS l
. To eliminate the possibility that a single electrical failure could prevent shutdown of the reactor, electrical separation in the areas of powe.r supplies, cable tray allocations, etc., is required.
in general, this separation provides for reduhdant means to provide boration and askeup to the RCS.
In addition local control stations enable the operators to maintain the plant
~
in a safe condition, assuming that the control room is inaccessible.
1 2.4 REACTOR MAKEUP CONTROL SYSTEM OPERATION The reactor makeup control consists of a group of instruments arranged to provide a manually pre-selected makeup composition to the charging pump suction header or the volume centrol tank. The makeup control functions are to maintain desired operating fluid inventory in the volume control tank and to adjust reactor coolant boron concentration for reactivity and shim control.
The control switches are located on the main control board along with the batch integrators and the flow controllers. Two switches are provided, one j
for Off/ Manual /Sorate/ Auto Makeup / Alternate Oilute/ Dilute and one for Stop/ Neutral / Start.
Automatic Makeun The automatic makeup mode of operation of the reactor makeup control provides dilute boric acid solution, preset to match the boron concentration in the Reactor Coolant System. The automatic makeup compensates for minor leakage of f
reactor coolant without causing significant changes in the coolant boron concentration. It operates on demand signals from the volume control tank level controller (LICA-112).
4 10 80260:10/022285
+-
a-e-
j4 i
Under nomal plant operating conditions, the mode selector switch is set in the " Automatic Makeup
- position and the boric acid and reactor makeup water j
flow controllers are set to give the same concentration of borated water as j
contained in the Reactor Coolant System. The mode selector switch aust be in j
the correct position and the control energized by prior manipuistion of the I
' Start" switch. A preset low level signal from the volume control tank level controller (LICA-112) causes the automatic makeup control action to start a f
selected reactor makeup water pump, start a boric acid transfer pump, open the makeup stop valve (FV-1100), makeup water flow control valve (FV-111A) and j
boric acid flow control valve (FV-110A). The flow c'ontrollers automatically set the boric acid and reactor makeup water flows to the present rates.
Makeup addition to the charging pump suction header causes the water level in l
I the volume control tank to rise. At a preset high level point, the reactor j
makeup water pump stops; the boric acid transfer pump stops; the reactor makeup water and boric acid flow control valves close; and the makeup stop l
valve closes. This operation may be teminated manually at any time by l
actuating the makeup stop.
The quantities of boric acid and reactor makeup water injected are totalized j
by the batch counters and the flow rates are recorded on strip recorders.
l Deviation alams for both boric acid and reactor makeup water are provided if -
flow rates deviate from set points.
i Dilute f
The ' Dilute
- mode of operation pemits the addition of a preselected quantity of reactor makeup water at a pre-selected flow rate to the Reactor Coolant f
The operator sets the mode selector switch to "Oilute", the reactor System.
makeup water flow controller set point to the desired flow rate, the reactor l
makeup water batch integrator to the desired quantity and actuates the makeup i
The start signal causes the makeup control to start a selected reacter start.
makeup water pump and open the makeup stop valve (FV-1118) to the volume I'
The control tank inlet and the makeup water flow control valve (FV-111A).
makeup water is t. died tt:mugh the volume control tank spray nozzle and Excessive rise of the through the tank to the charging pump suction header.
11 80260:10/022285 l
. _,.,, -, ~.. - - - - - - - - -
dl' t
volume control tank water level is presented by automatic actuation of a i
i-three-way diversion valve (by the tank level controller), which diverts the l
When the pre-set reactor coolant letdown flow to the recycle holdup tanks.
quantity of reactor makeup water has been added, the batch integrator causes the reactor makeup water pump to stop and the reactor makeup water " control
{
This valve and reactor make up water stop valve to the VCT inlet to close.~
1 operation may be terminated manually at any time by actuating this makeup stop.
j l
I Alternate Dilute i
The alternate dilute mode is similar to the dilute mode except a portion of the dilution water flows directly to the charging pump suction and a portion j
j-
. flows into the volume control tank via the spray nozzle and then flows to the charging pump suction.
The operator sets the mode selector switch to " Alternate Dilute", the reactor i
makeup water flow controller set point to the desired flow rate, the makeup l
water batch integrator to the desired quantity and actuates the askeup start.
l The start signal causes the makeup control action to start a selected reactor l
makeup water pump and opens the makeup stop valve to the volume control tank and the makeup stop valve to the charging pump suction header and the reactor j
makeup water control valve. Reactor makeup water is simultaneously added to This the volume control tank and to the charging pump suction header.
minimizes the delay in having to dilute the volume control tank before the RCS
[
Excess water level in the volume control tank is prevented by can be diluted.
automatic actuation of the volume control tank level controller, which diverts When the preset k
the reactor coolant letdown flow to the recycle holdup tanks.
quantity of reactor makeup water has been added, the batch integrator causes the reactor makeup water pump to stop and the primary makeup water control valve and the reactor makeup stop valves to close. This operation may be terisinated manually at any time by actuating the makeup stop.
'l 12 8026g:10/022285
.n..-....m-.
t i
AarAit The borate mode of operation permits the addition of a pre-selected quantity l
l of concentrated boric acid solution at a pre-selected flow rate to the Reactor Coolant System. The operator sets the mode selector switch to *Bo mte' the concentrated boric acid flow controller set point to the desired flow rate, l
f the concentrated boric acid batch integrator to the desired quantity and I
actuates the makeup start. Actuating th.e. start switch opens the makeup stop l
valve (FV-1108) to the charging pump suction and the boric acid control valve I
The concentrated f
(FV-110A) and starts the selected boric acid transfer pump.
boric acid is added to the charging pump suction header. The total quantity j
added in most cases will be so sen11 that it will have only a minor offect on the volume control tank level. When the preset quantity of concentrated boric j
acid solution has been added, the batch integrator causes the boric acid transfer pump to stop and the concentrated boric acid control valve and the f
makeup stop valve to close. This operation may be terminated manually at any i
time by actuating the makeup stop.
i Makeuo Stop By actuating the makeup stop, the operator can terminate the makeup operation in any of the four modes of operation.
5tpJLa.1 The manual mode of operation permits the addition of a preselected quantity of boric acid solution at a preselected flow rate to the refueling water storage tank or through the temporary (flanged) connection to other items of l
equipment. While in the manual mode of operation, automatic makeup to the Reactor Coolant System is precluded. The discharge flow path must be prepared by opening manual valves in the desired path.
The operator then sets the mode selector switch to " Manual", the boric acid and makeup water flow controllers to the desired flow rates, the boric acid and makeup water batch integrators to the desired quantities and actuates the
(
80260:10/022285 13
4 makeup start switch. Actuating the start switch activates the boric acid flow control valve (FV-110A) and makeup water flow control valve (FV-111 A) and starts the preselected reactor makeup water pump and boric acid transfer pump.
1 i
When the present quantities of boric acid and reactor makeup water 1mave been j
j added, the pumps stop and the boric acid and askaup water flow control valves close. This operation may be stopped manually by actuating the makeup stop i
switch.
i If either batch integrator is satisfied before the other has recorded its j
required total, the pump and valve associated with the integrater which has been satisfied will terminate flow. The flow controlled by the other The horic acid integrator will continue until that integrator is satisfied.
flow rate should always be set slightly higher than the required mixture rate, to insure that boric acid flow is terminated and the lines are flushed I
by reactor makeup water.
Alare Functions i
The reactor makeup control has been provided with alarm functions to call the i
i f
operator's attention to the following conditions:
Deviation of total makeup water flow from control set point.
l a.
r 1
Deviation of concentrated boric acid flow rate from control set point.
I b.
i l
4 4
80260:10/022205 14 1
~
~
4 i
2.5 CHEMICAL ANO VOLUME CONTROL SYSTEM OPERATION l, '
j 2.5.1 PLANT STARTUP a
Plant startup is defined as the operations which bring the reactor plant from i
the cold shutdown condition to nonnel, no-load operating temperature and 1
i pressure, and subsequently to full-power operation.
5 1
The charging pumps initir.11y fill and pressurite the reactor coolant system.
l During filling, makeup water is drawn from the reactor makeup water storage f
tank and. blended, using the Reactor Makeup Control System, with boric acid, to provide makeup water to the prevailing reactor coolant system boron l
The Reactor Coolant System is vented via the reactor vessel l
concentration.
head. The pressurizer is vented separately to the pressurizer relief tank.
i Following the venting operation, a letdown flow path is established by opening l
the letdown valve and the low-pressure letdown control valve. The pressurizer heaters are energized to start increasing the pressurizer temperature.
j l
Cleanup via the residual heat removal loop is terminated. The charging pumps are then employed to increase the reactor coolant system pressure (water solid l
at this time). The manual throttle valves in the seal water supply lines are f
l set to provide labyrinth and No. 1 seal flow (-8 gps / pump).
The rate of increase of system pressure is controlled by manual operation of l
l If the low-pressure letdown valve and regulating the charging pump flow.
desired, the low-pressure letdown valve may be set in " Auto" to maintain a l
pressure of about 400 psig in the letdown system downstream of the orifices; l
pressurization is then controlled by the charging system.
When the reactor coolant system pressure has reach =400 psig, residual heat l
removal is terminated, and the AP across the No. 1 seal leakoff is checked.
If in order, a check is made to ensure that seal-water in.jection flow is l
The reactor coolant pumps are started sequentially.
If chemical l
adequate.
The treatment such as hydrazine addition is required, performed at this time.
mixed-bed domineralizers are bypassed during chemical addition. The Reactor l
l i
80260:10/030685 15
i i
Coolant system will heat up due to the reactor coolant pump heat input and l
l.
l residual heat addition; hence excess coolant will accumulate in the volume control tank. The volume control tank level H ses and the nitrogen cover gas l
is espelled to the waste processing system.
As soon as high level is reached in the volume control tank, the nitrogen i
supply is secured and the hydrogen makeup valve is brought into oporttion.
l DuM ng this operation the volume control, tank pressure is maintained at =15 The volume psig by the pressure control valve in the gaseous vent line.
l control tank level is allowed to decrease to normal W manually diverting the 1etdown flow to the recycle holdup tanks. This operation establishes the l
hydrogen overpressure in the volume control tank.
At this Heatup is continued until a temperature of about 250'F is achieved.
point, the pressuM zer heaters are employed to draw a steam bubble in the pressurizer. The low-presture letdown control valve is now set in ' Auto" to maintain about 350 psig downstream of the letdown oHfices and charging-pump flow is controlled manually to obtain normal water level (no load) in the pressurizer. The charging pump can be placed in ' Auto' following attainment i
of normal water level. As heatup proceeds it will be necessary to provide extra letdown flow capability is provided by opening selected orifice isolation valves. This requirement will be dictated by the regenerative heat exchanger (a maximum of 380*F is allowed at the outlet from the heat exchanger upstream of the letdown orifices), and the rate of expansion of the coolant due to heatup as reflected by pressuMzer level. The excess letdown heat exchanger may be employed as the reactor coolant temperature approaches T, 1
no-load to accelerate the heatup phase. The pressuM zer heaters are employed e
[
periodically duMng heatup to ensure a temperature difference between the coolant in the pressurizer and the reactor coolant system lines of at least 50*F but not more than 200'F.
Following chemical analysis to establish that water quality, boron concentration and hydrogen concentration are within specification, criticality is achieved by appropriate rod withdrawal; subsequent reduction of boron s
I v.
80260:10/022285-16 b
Further adjustments in boron concentration by dilution will be required.
concentration by operation of the Reactor Makeup Control System to establish preferred contro1 1 roup rod positions and to compensate for menon buildup will also be necessary.
l Following attainment of full power, the letdown orifices are set for normal 1etdown.
l During the heatup phase it should not be necessary to adjust the seal water injection valves; however, some adjustment of the charging line control valve may be required to maintain the required seal injection flow.
k 2.5.2 NORMAL OPERATION Normal operation includes operation at steady power (base load) level, load k
follow operation and het standby.
anse Load D
At a constant power level, the rates of charging ~and letdown are dictated by the requirements for seal water to the reactor coolant pumps and the normal One charging pump is employed and purification of the Reactor Coolant System.
is controlled automatically from pressurizer level. The only adjustments in boron concentration necessary are those to compensate for core burnup.
y L'
I These adjustments are made at infrequent intervals (= twice per week) to
(
I Rapid maintain the rod control groups within their allowable '11mit'.
h variations in power demand will be accommodated automatically by control rod If variations in power level occur, and the new power level is movement.
sustained for long periods, some adjustment in boron concentration may be necessary to ensure preservation of shutdown margin.
x I
During normal operation the letdown flow is 75 gym and one mixed-bed Reactor coolant samples are taken at frequent l
domineralizer is in service.
intervals (= once per shift) to check boron concentration, water quality, l
1
.17 80260:10/022285
-w aay-a w
.-ea,
---c-----.
m
I l
pH, and activity level. The charging-pump speed control or flow control valve position maintains the pressuriser water level at the setpoint programmed for a prevailing reactor coolant average temperature. During operation at constant power the Reactor Makeup Control System is set in ' Auto" to, provide leakage makeup at prevailing reactor coolant system boron concentration.
Makeup is initiated automatically if the volume control tan level falls to the low-level setpoint.
j Doerator Action for Load Follow When a change in plant lead occurs, the control system will position the control rod banks in accordance with the load dependent program for reactor coolant temperature. The Boron Thermal Regeneration System is then employed to effect the required boration concentration change in the Reactor Coolant System to accommodate the reactivity transients which occur as a result of e
load changes.
2.5.3 PLANT SHUTOOWN Not Shutdown If required, for periods of maintenance, or following spurious reactor trips, the reactor can be held subcritical, but with the capability to return to full
,a power within the period of time it takes to withdraw control rods. During this hot shutdown the average temperature is maintained at no-load T,,, by initially steam dumping to provide residual heat removal, or at later stages by running reactor coolant pumps to maintain system temperature.
l Following shutdown, tenon buildup occurs and increases the degree of shutdown; i.e., initially, all control rods are inserted and the core is maintained at a minimum of 2 percent Ak/k subcritical. The effect cf xenon buildup is to increase this value to a maximum of about 3 percent Ak/k at about nine hours following shutdown.
i i
l]
80260:10/062085 18 w
T
~
w
--++
=
==zzzz;
- - - ^
?.
If rapid recovery is required, dilution of the system may be performed to l -
counteract this menon buildup. A shutdown group of rods must be Mt'; drawn during dilution and frequent checks made on critical red position.
l l
Plant Shutdown l
Plant shutdown is defined as the operations which bring the reactor plant from l
normal operating temperature and pressure to cold shutdown for maintenance or l
refueling.
I Sefore initiating a cold shutdown, the volume control tank overpressure is This reduced to lower the Reactor Coolant System hydrogen gas concentration.
{
requirement will also apply to cold shutdowns for refueling purposes.
Also, the reactor coolant boron concentration is increased to the cold The operator sets the reactor makeup control to "Sorate",
i shutdown value.
selects the volume of concentrated boric acid solution necessary to perform the boration, saspies the reactor coolant to verify that the concentration is,
correct, and sets the reactor makeup control for leakage askeup at the shutdown reactor coolant boron concentration.
Contraction of the coolant during cooldown of the Reactor Coolant System results in actuation of the pressurizer level control to maintain normal
[
pressurizer water level. The charging flow is increased, relative to letdown The volume flow, and results in a decreasing volume control tank level.
control tank level controller automatically initiates makeup to maintain the inventory.
f After the Residual Heat Removal System is place in service and the reactor coolant pumps are shutdown, further cooling and depressurization of the pressuriter fluids are accomplished by charging through the auxiliary spray connection.-
i e
~
t 80260:1D/022285 39 j
i 1
-.. - ~
- =. - -
1 l
1 If._ required, the mixed-bed domineralizer cad gas stripping in the v31ume l'
control tank can be operated at maximum letdown in advance of a planned i
shutdown. Domineralization of ionic radioactive impurities and stripping f
fission gases reduce the reactor coolant activity level sufficiently to permit i.
personnel access for refueling or maintenance operations.
I l
2.5.4 ABNORMAL OPERATION
(
j Loss of Nor1 mal Letdown i
l e -
If the normal letdown path is lost, an alternate letdown path through the l
excess letdown heat exchanger is available. The excess letdown flow is l
sufficient to. allow the pressurizer level to be maintained while seal
. injection water flow to each reactor coolant pump is continued.
j I
Reactor Coolant System Leak i
The CVCS is capable of making up for a small RCS leak of approximately 130 spa p
i.
using one centrifugal charging pump while maintaining seal injection flow to the reactor coolant pumps. This capability includes allowance for a minimum 4
l RCS cooldown contraction, and assumes the letdown line will be isolated on low pressurizer level (LV-45g, 460).
i l
l i
l-k I
L 1
20260:10/022285 20 9
- - _ = - - - -. -..
l A
3.0 ANALYSIS 3.1 INITIATING EVENTS 1
The possible events which could initiate a boron dilution accident wre L
determined using a failure modes and effects analysis (FMEA) of the Chemical i
j and Volume Control system. Each component of the system was considered to
~
f determine the consequences of its failure For those components whose failure -
1 The results could lead to boron dilution, associated alems wre,dentified.
I l
of the FMEA. appear in Appendix A.
Each failure which was identified as a f
possible boron dilution initiator was examined to determine if, because of administrative controls or similar precautions, it would require a double l
failure to lead to boron dilution. Where double failures, rather than single, j
were necessary to cause an event, the probability of the event becomes much Table 3-1 l
lower than the single failure events and is, thus, bounded by them.
lists all of the possible initiators which were identified by the FMEA, the modes in which they apply, and any procedure which would make the initiators p
A final list of four single failure initiators was require double failures.
I produced and is shown in Table 3-2.
Both i
Next, the f requencies of these four initiators were calculated.
4 mechanical failures and human errors were considered to determine the l
The following paragraphs explain how each of the frequencies was frequencies.
l calculated. All human error data was taken from NURES/CR-1278.
t DENINERALIIER OUTLET ISOLATION VALVE OPEN DURING RESIN FLUSHING 3.1.1 1
The main contribution to this event is failure of the operator to close the l
isolation valve, as stated in the procedure. The valve could also transfer f
l open due to mechanical failure, but this is less likely than the human error.
The following assumptions were made in the analysis:
)
i P
l 4
h 21 p.
80260:10/062085
'q.
The procedure involved includes more than 10 steps.
a) i b) Each Domineralizer is flushed once per year.
i c) One hour is required for each flushing operation.
f using these assumptions, a frequency of.03 per reactor year,was calculated, i
based on the following data:
human error (omit step in written procedure).012 (Table 15-3, Reference 4) human error (incorrect valve operation)
.003 (Table 14-1, Reference 4) frequency of domineralizer flush (per reactor year) 1 This procedure is normally performed in cold shutdown and 805 of the initiating events were assumed to occur in mode 5A, and 205 in mode 58.
I However, administrative controls prevent the occurance of a boron dilution incident from this event during mode 58.
l VALVE 226 OPEN FOLLOWING BTR5 DEMINERALIZER FLUSHING OPERATION 3.1.2 This event also involves a human error or a mechanical failure. The operator could fail to close valve 226 following domineralizer flushing. This is very similar to the error mentioned above in that it is an omission of a step in a long procedure that is used about once per year. This procedure was also assumed to be performed twice per year and thus the frequency of this initiating event was also calculated as.03 per reactor year.
3.1.3 FAILURE TO SECURE CHEMICAL ADDITION f
This event could be caused by a human error or by a double mechanical failure of valves 176 and 181. The double failure has a negligibly low probability.
The human error involves leavint out a step in a short procedure.
It was estimated that chemical addition is used about 25 times per year in modes 3 and 4 and once per year in modes 5A and 58 each.
80260:10/062005 22
Thus the frequency of this initiating event a s calculated as 0.12 per reactor year in modes 3 and 4 and 0.0094 per reactor year in modes 5A and 58. The probability of failure to secure a specific chemical addition operation was
~3 calculated as.0047 (3.7 x 10 for omission of a step (ref,erence 4,' Table 15-3) and.001 for incorrect valve operation (Table 14-1).
3.1.4 VALVE FV-110A FAILS CLOSED DURING,MAKE-UP This event involves mechanical failure of the flow control valve from the Boric Acid Storage Tank. The hourly failure rate for this type of valve is 1.5 x 104 (Reference 3)..
The following conservative estimates of times spent in modes were assumed:
I mode 3
- 20 days /yr mode 4
- 40 days /yr mode 5 (filled) - 60 days /yr mode 5 (drained) - 10 days /yr Reactor make.up from this path is isolated by locked closed valves during mode 58 and mode 6 operation.
The resulting initiating event frequencies were calculated:
~3 modes 3, 4
- 2.2 x 10 /yr
-3 mode 5 (filled) - 2.2 x 10 /yr Note that modes 3 and 4 are combined because their similar volumes and boron concentration allow one event tree to cover both modes.
80260:10/062085 23
_- --.. _. =
f 3.2 OPERATOR RESPONSE TIMES Rest of the boron dilution events require operator action to prevent unplanned
[
criticality. The operator may either diagnose the problem and correct it or In this may react to an alars and follow the corresponding procedure.
analysis operator response was imodeled in response to either the high flux alarm or to the high volume control tank level alars. Conservatively 7no credit was taken for operator response to,, increasing source range counts or to boronmeter indications, however for very slow dilutions (greater than 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br />) the possibility of detecting the event during the daily verification of shutdown margin was considered. The time for response to the alarm is limited to the time from alare actuation to unplanned criticality. Appendix 8 contains the derivation of the equations and the calculation of the available Calculations are shown for the BA Flow Deviation Alars, the Volume times.
An alars Control Tank High Level Alars, and the High Flux at Shutdown Alarm.
actuates more than 15 minutes before criticality for each initiator and in This shows that the design meets the 15 minute time twouirement of each mode.
the Standard Review Plan.
I l
t l
l 80260:10/061885 24 lf i
l i
3.3 EVENT TREES The first l
The responses to each initiating event were modeled in event trees.
two events had similar sequences of responses; therefore, the same event tree I
. structure was used for all three. The third and fourth had same different l
responses, which required separate event trees. Three modes were considered:
i
~
i
- 1) modes 3 and 4. combined f
- 2) mode 5. filled (mode 5A)
- 3) mode 5. drained (mode 58) f Figure 3-1 shows the event tree for initiators 1 and 2.
The top events of the l
tree are explained below:
i 4
ET1 or ET2 - 1eproper CVCS or STRS Domineralizer Flushing TLA - Volume Control Tank Level l
HFA - High Flux Alarm DAI - Operator' diagnoses problem and isolates cause These initiators result in an inflow of dilute water into the letdown flow
)
i upstream of the Volume Control tank.
In response to these events, the VCT level will increase and the operator will receive VCT high level indications (i.e. VCT high level alarm and/or indication of flow diversion to the recycle holdup tank). If the operator does not receive or respond to the VCT level
[
indications the high flux alarm will actuate and the operator will terminate the dilution event and restore shutdown margin.
Figure 3-2 illustrates the event tree for the failure to secure chemical
[
addition event. The top events of the tree are as follows:
ET3 - Initiating event - Failure to secure chemical addition
{
NFA - High flux at shutdown alars actuates DAI - Operator diagnoses dilution event and isolates.
s-9 80260:10/061885 25
Thie ;nitiating event (failure to secure chemical addition) results in en Because of the long extremely slow dilution of the reactor coolant system.
l time involved in the dilution event, operator diagnosis of the boren dilution event is likely during the daily shutdown margin verification even if the high flux alarm fails to operate.
The event tree which represents the fourth initiator (the failure clohd of FV-110A) appears in Figure 3-3.
The top,pvents of the tree are as follows:
ET4 - Initiating event - Failure closed of FV-1104 FA - Flow deviation alarm actuates MUI - Makeup is isolated either automatically or by operator I
HFA - High flux at shutdown alarm actuates DAI - Operator. diagnoses problem and isolates source This event would cause almost immediate Soric Acid and Makeup flow deviation I
i alarms and automatic isolation of makeup. If the flow deviation alarm or automatic isolation of makeup fails then the high flux alarm shculd annunciate l
~
and the operator would respond by isolating charging and initiating emergency 1
boration.
I 1
I
(
I a
26 00260:10/030885
~
l'
-i--..-
.,,,,.,r
3.4 FAILURE PROSABILITIES FOR TOP EVENTS The probabilities of failure of the top events or nodes, of the event trees were calculated using the data in Appendix C.
Table 3-3 lists the values for each node of each tree and summarizes the results of this section.
3.4.1 HIGH FLUX ALARM (HFA)
Technica1 Specifications require that at least one source range detector be operable during shutdown modes 3. 4 and 5, with a channel check performed Based on the every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and an analog channel operational test monthly.
I above and data from Appendix C a failure probability of 3.g x 10 was calculated for this alarm.
3.4.2 VOLUME CONTROL TANK LEVEL ALARM (TLA)
(
The volume control tank has two level detectors, however the high level alare is actuated by only one detector. Credit for the redundant detector was taken only for facilitating in the detection of failures of the high level alam detector and associated instrumentation. A total failure probability of g.1 x 10 was calculated for this alam (Appendix C).
-3 3.4.3 FLOW DEVIATION ALARM (FDA)
Failure closed of the boric acid flow control valve will result in immediate control room annunciation via the boric acid flow deviation alarm.
Additionally the total makeup flow deviation alarm will actuate if the boric r
F acid flowrate is greater than 8 gym as is nomally the case. The probability
~3 of neither alarm actuating was calculated as 7.3 x 10 3.4.4 MAKEUP ISOLATED (MUI)
This event tree node is addressed in the boric acid flow valve failed closed initiating event (event tree 4). If a flow deviation alarm is actuated.
1 M
s e
=
ll.
q FV-1108 and FV-1118 automatically close to isolate makeup flow. A failure probability of 4.0 x 10-2 was calculated for this event tree node.
3.4.5 DIAGNOSE AND ISOLATE N00E (OAI)
Success of,this action is addressed by all four boron dilution event. trees and is affected by a number of variables including:
1.
the alams which have been annunciated 2.
the time available between alarm annunciation and loss of shutdown margin.
s For the high flux alars, it is assumed that all operators are familiar with the iemediate actions to the alarm which will include isolation of the normal charging flow and actuation of emergency boration. Additionally it is assumed l'
that the Abnormal Operating Instructions will be available in the control room and that the operators will be trained to refer to the appropriate procedure in the event of a high flow alarm. Under these conditions the lower bound human error probabilities of figure 12-4 of reference 4 are applicable.
If only high volume control tank level indications are received in the control room, the operator response may be impaired because of misdiagnosis of the
)
cause of the alarm. For some events it is possible that the operator will I
attribute the high VCT level to expansion of reactor coolant or to makeup operations.
For these reasons the upper bound curve of figure 12-4 of reference 4 was used to estimate the response to this alarm. Additionally a f ailure probability of.05 was added to account for misdiagnosis errtrs if greater than 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> was available for detection and.2 was added if b9 tween 30 and 60 minutes was available.
If both the VCT level alam and the high flux alarm are received in the control room the product of the human error i
probabilities is used to estimate the total human error probabilities since the two alarms would result in event termination by different methods.
I 1
i 5
J I
d ll 80260:10/030885 28
For the f ailure to secure chemical addition evont tree, operator diagnosis and event termination is considered even in the absence of a positive alare due to In this case the daily the long time required for loss of shutdown margin.
shutdown margin check may discover the boron dilution event and result in If the time to operator action to recover the required shutdown margin.
criticality was greater than 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br /> then the human error probability (NEP)
For modes SA and 58 the, HEP was calculated as 1,0- 0.7 was estimated as.01.
[(time to criticality)/24 hour) +.01.
J s
1 j.
i
- \\
=
l 9
29 l
80260:10/030885
l TABLE 3-1 POSSISLE INITIATING EVENTS comment: -
1xani Bedst 4.
Manual valve 183 left open All Double failure because of follwing test or maintenance independent safety j
verification i
{
2.
Domineralizer outlet 3,4,5 In mode 6, domineralizers are isolated by locked valves isolation valve open during resin flushing 3.
Failure to secure chemical 3, 4, 5 Credible event, but extremely slow dilution addition (valves 176 and 181 open)
L 4.
valve FV-110A fails closed 3,4,5 Makeup isolated during mode 6 during makeup, causing loss of boric acid flow 5.
Valve 226 left open following 3,4,5 STRS isolated during mode 6 domineralizer flushing f
6.
SA Storage Tank isolation All Valve is locked open and has independent safety f
valve closed during makeup verification following maintenance hk e
30 80260:10/030685 m
1
TAsLE 3-2 INITIATING EVENTS INVOLVING SINGLE FAILURES C
)
amtsi LVsnt
- 3. 4. 5A 1.
Desineralizer outlet isolation valve 1
open during resin flushing
)
3, 4, 5A 2.
Valve 226 open during BTRS operation 3,4,5 3.
Failure to secure chemical addition
- 3. 4. 5A 4.
Valve FV-110A fails closed during makeup t
e d
i I
l l
i i
k n
h l
sor6o:in/osisms 31
TABLE 3-3 TOP EVENT FAILURE PROSABILITIES Event Tree M
Node 3.4 M
Mode 5b 1
ET) 2.4[-3
.024 g
TLA 9.1E-3 9.1E-3 HFA 3.9E-4 7 9E-4 DAI (loth alarms) 5E-6 SE-6 DAI (HFA only) 1E-4 1E-4 DAI (TLA only) 5.0E-2
.05 2
ET2 2.4E-3
.024 TLA 9.1E-3 9.1E-3 NFA 3.9E-4 3.9E-4 l
DAI (TLA only)
.06
.06 DAI (HFA only) 4E-4 4E-4
/
l DAI (Both alarms) 2.4E-5 2.4E-5 s
l 3
ET3 0.12 9.4E-3 9.4E-3 l
HFA 3.9E-4 3.9E-4 3.9E-4 DAI (HFA) 1E-6 IE-6 1E-6 DAI (No alarm)
.01
.01
.31 4
ET4 3.3E-3' 3'.3E-3 FA 7.3E-3 7.3E-3 MUI 4.0E-3 4.0E-3 HFA 3.9E-4 3.9E-4 DAI 2E-3 2E-3 80260:10/061885 32
_.... _.. _ _.... _ _. _.. _.. _... _. -... _. _ ~......... _ _ _.. _.. -.. _..
H f,
i; P
t!
J t
t F
IFA MI I SUCCESS TLA ETg r
aLa i
i 3 success E
4LM
=
2 s success m
Stm i
I TLW g
3 l
LSM = toss Of Shutdown Margin Possible FIGUllE 3 - 1
?
EMMT TEE 1 - CVCS Demineralizer Outlet Valve Open During Flushing EKMT TKE 2 - Valve 226 Open following BTRS Deminerallrer Flushing
+'----w-
=-~d m
a A
w A2.
s
,na a
~
-.6 aa e
O S
j E.
E a
em 08 PD T J
i misa 4
e 1
I g
1 as=
'S
+
id e
m
==
==.
E=U
(
3.
t n
a i
w C
s O
=
eum S
me b
g 6
I 09 WN-4 i
E 1
W w
1 l
M
\\
J di I
34 I
ET4 FA MDI WA 948 1 Suett90 t success s
=
3 LSn 4 Len i
5 3UCCESS i
i a ten
=
=
T LSM i
ue FIGURE 3 - 3 i
FV-110A Falls closed During Makeup EVENT TREE 4 -
I i
l l
I
4.0 RESULTS The event trees were quantified using the frequencies and failure probabilities presented in Section 3.
The results of each tree are 31sted in Table 4-1, along with the total frequency of unplanned criticality.
A total frequency of 4.0 x 10 per reactor year is calculated for th 4
occurrence of a loss of shutdown margin d've to boron'dilutiot) events. The frequency of loss of shutdown margin events was initially don'.'nated by domineralizer flush valve misalignments, particularly in mode 58. Vogtle administrative procedures have since been modified to require isolation of the dilute water supply to the primary and BTRS domineralizers prior to proceeding Additionally this analysis assumed that the domineralizer flush to mode 58.
procedures would include strict administrative controls (including independent These
/
verification of the valve lineup) to prevent an inadvertent dilution.
administrative controls and the redundancy of alarms accounts for the low frequency of loss of shutdown margin events.
4 i
l t
i i
l l
l I
1 l
i 80260:10/062085 36
-j.
t TABLE 4-1 i
RESULTS Frequency of Unplanned LL.
Egit criticality (Per Year) 5 Contribution 1
3, 4 7.01-8 2
5a-7.0E-7 17 5b 2
3, 4 1.3E-7 3
i Sa 1.3E-6 32 i
\\
}
Sb
)
i i
3 3, 4 5.9E-7 14 5a 4.6E-8 1
Sb 1.1E-6 27 4
3, 4 8.9E-8 2
Sa 8.9E-8 2
l Sb i
i Total 4.0E-6 9
I l
1 t
l l
l 1
E 1
l 80260:10/061885 37
l
!=
l 1.
\\
5.0 CONCLUSION
S i
As was mentioned in Section 3.2. the times between an alarm and unplanned criticality for all initiating events in all modes are greater than 35 This allows sufficient time for the operator to respond to the minutes.
It also shows that the 15 minute minimum requirement specified in the j
events.
Standard Review Plan is met for all events.
4 1
~
t 1
l i
i 1
1 I
i k
I 80260:10/062085 38
REFERENCES 1.
Letter from Tom Gerlowski to Sherrie Erwin, F50/55-GAE-3590, January 8, 1985.
2.
Personal Communication with Arto Cinar to Sechtel, November 5,1984.
3.
NURE6/CR-2770, " Common cause Fault Rates For Valves", Steverson, J. A. and Atwood, C.
L., February 1983.
4.
NURES/CR-1278, Handbook of Human Reliability Analysis with emphasis on Nuclear Power Plant Applications Final Report, A. D. Swain and H. E.
?
' Guttman, June 1983.
j 4
5.
Millstone Unit 3 Probabilistic Safety Study. August, 1983.
l 6.
Annunciator Response Procedures for ALS-07 Rev. 0,17007-1, Georgia l
Power Vogtle Electric Generating Plant.
l I
j-1 i
}
{
80260:1D/061885 39
.. -.. ~.....
y
'i APPENelK A
,e FAILUREMootAneEFFECTSANAliSISCHEMICALANSVolullECONTROLSYSTEM ACTIVE COMPONENTS - NORMAL PLANT OPERATION AND SAFE SNU190MN CVCS Operation Effect on System component Fa11ere Mode Function Deeration and Shutdown Additional Inferentien**
l 1.
Air diaphragm a.
Falls open a.
Charging and Volume a.
Failure redeces Control - letdown redundancy, of operated globe flow.
providing letdown valve LV45g flow isolation to (LV460 analogous) protect PRZ heaters from uncovering at low water level in PRZ. No offett on system operation.
Alternate 1selatten s
valve (LV-460) provides backup i
w letdown flow L
isolatten.
- b. ' Falls closed b.
Charging and Volume 6.
Fallure blocks a Control - letdown normal letdown flow.
flow to VCT. Minimum letdown flow re-quirements for boratten of RCS of safe shutdown concentration level I
may be met by
,t estab11shing letdown flow through alternate, excess letdown flow path.
If the alternate, excess letdown flow l
~.
.____.__..._._._._....__.___e APPENDIX A (Cont)
CVCS operation Effect on System empement Failure node Function emeratten and shutdeus peserts path to.VCT.ts not l
available due to.
common mode fa11ere (less of instrument air supply) affectlag the opening operaties of isolatica valves in each flew path, the plant operater can herate the RCS to f
a safe shutdeva concentratten level l
withest letdeus flow l
by taktag advantage l
ef the steam space i
available la th PRI. Letdeva can s
aise he provided from the reacter vessel head.
2.
Air diaphragm a.
Falls open.
a.
Charging and volume A.
Fa11ere prevents Centrol - letdown isolatten of normet operated globe flow.
letdeua flew thre* A l
valve M-41438 regenerative heat (014sc and exchanger. No effect 814gA analogous) safe shutdeum operetten.
Contatament isolatten valve (NV-8152 er 4
8160) may he remotely tissed from the CS te 1selate letdown flow through heat exchanger.
+
i i
80260:19/030685 u.
1 i
APPEN0!X A (Cont)
~
CVCS Operation Effect on System Failure Mode Function Operation and Shutdown Additional Information**
oggpnent b.
Falls closed.
b.
Charging and Volume b.
Failure blocks normal Control - letdown letdown flow to VCT.
Normal letdown flow flow.
to VCT may be maintained by opening alternate letdown orifice isolation valve NY-814gt.
Minimun letdown flow requirements for boration of RCS to safe shutdown con-
\\'
centration level may l
be met by opening 1etdown orifice 1 solation, valve t
HV-814gA or 814gC.
If casuna mode failure (loss of instrument air) prevents opening of these valves also i
prevents establishing alternate flow through excess letdown flow path, 4
plant operator can borate the RCS to a safe shutdown i
concentration level without letdown flow i
by taking advantage of steam space available in PRZ.
Letdown can also be provided from the reactor vessel head.
1 i
annn 1R/022285 l
APPENelX A (Cont) 3 CVCS Operation Effect on System t
2
. Component Failure Mode Function -
Operation and Shutdown Additional Infovisation**
3.
Air diaphragm a.
Falls closed.
a.
Charging and Volume a.
Same offact on system operated globe-Control - letdown operation as that
- valve MV-0152 flow.
stated for ites No. 1 failure mode ' Falls (0160 analogous) closed".
b.
Falls open.
b.
Charging and Volume b.
Failure has no etfact Control - letdown on CVCS operation
- flow, during normal plant operation. However, under accident con-ditions requiring i
containment isola-tlon, fallure reduces the redundancy of providing isolation of norinal letdown i
I line.
i l
~
~
4.
Air diaphragm a.
Falls open.
a.
Baron Concentration a.
Failure inhibits use operated globe Control - boron of STRS for load valve TV-3elt thermal regeneration follow operation i
(boration).
(boration) due to low temperature of let-l down flow entering SIRS dominera11rers.
Alternate horation of r
reactor coolant is 1
possible using RNCS i
of CVCS. No effact on operation to bring i
i reactor to safe shutdown condition.
I t
s an,an.inins,sas
.._. _ _ _.. _ _. _... _... _7.. - _.7. -
APPEN0lX A (Cont)
CVCS Speratten Effect en system anement Fa11mre Meet Functles Baeraties and shutdensi Addittenal Infomatten**
b.
Falls closed.
b.
toren Concentration b.
Failure inhibits use Centrol - beren of STR$ for lead thermal regeneratten follow operetten I
(boratten).
(boratten) due to less of temperature control of letdown flow entering STRS deelnera11rers.
Failure aise blocks normal letdeun flew 1
-te VCT when STR$ is not being used for t
lead follow. Minimms letdown flew re-quirements for i
beratten SCS to safe s
l shutdown concentratten 1evel may be met as i
stated for of f act en system operatten for item No. 1. failure
" Falls closed".
- i. Air diaphragm a.
Falls open.
a.
Charging and Vclose a.
Failure prevents c
Centrol - letdeun control of pressure operated globe flow.
to prevent flashing valve pV-131 of letdeun flew in l
letdown heat ex-
~
changer and aise allows high pressure fluid to mix bed de-minera11rers. Relief
' valve p5V-811g opens i
in deelneralizer line to release pressure
===e 1. s a ensas at
L l
ApptNORE A (Cent)
CVCS Operatten-Effett en System Campaamat failure lhet I'--^ttaa emeratten and shutdem Mditteaal Infomatten**
to VCT and valve (TC-12g) changes positten to divert flow to VCT.
Soration of RCS to safe -~.
shutdown concentratten i
level is possible with valve fa111eg open.
1 I
b.
Falls closed.
b.
Charging and Volume b.
Same effect.en system Centrol - letdown operation as that for flow.
item No. 1, fa11ere l
mode "Falled closed".
l 6.
Air diaphragm a.
Fatis open for a.
Charging and Volume a.
Letdown flew bypassed operated three-flew only to Centrol - letdeun from flowing to mixed way valve TV-12g VCT.
flow.
bed dominera11rers and SIRS. Failure prevents ionic puriflestion of letdeun flew and l
prevents operatten of
(
81R5. Geratten of 3
.RCS to safe shutdeun concentratten level 1s possible with valve fa111ag open I
for flew only in VCT.
b.
Falls open for b.
Charging and Volume b.
Contineens letdeun to i
flew only to Centrol - letdown mixed bed de-31N94 bed IISW sinerailrers and BTRS.
Fa11ere prevents deminera11rer.
automatic 1selatten of 90760:10/030685
._m.
i APPEN0!K A (Cont)
CVCS Operatlon Effect on System I
. "easonent Fa11ere Mode Function Operation and Shutdown Addittenal Infomatten**
mixed bed de-mineralizers and STRS under condition of e
high letdown flew ten-peratures. Doratten of RCS to shutdown concentration level is possible with valve fa111eg open for flow only to dominera11rers.
1.
Air diaphrage a.
Falls closed, a.
Charging and volume a.
Failure prevents use Control - excess of the excess letdeun operated globe letdown flow.
11ae of the CVCS as an valve NV-8153 alternate path that
'(0154 analogous) may be used for r
1stdown flew centrol.
4 t
i b.
Falls open.
b.
Charging and Volume b.
Failure reduces re-Control - excess dondancy of providing letdown flow.
excess letdown flew 1selatten during t
nonnal plant l
operation and for plant startup. No effact en system operation.
I i
i
~.
annnan/022285 i
i:
t' APPEN91X A (Cont) l CVCS Operation Effect on System I,
'onsonent Failure Mode Function Operation and Shutdown Addittenal Information**
1.
Air diaphragm a.
Falls closed.
a.
Charging and Volume a.
Failure prevents use operated globe Control - excess of excess letdown line l
. valve NY-123 letdown flow.
of the CVCS as an al-ternate path that may be used for letdown flow control.
I b.
Falls open.
h.
Charging and Volume b.
Failure prevents Control - excess manual adjustment letdown flow.
from control board i
L (CS) of RCS system pressure downstream of excess letdown heat exchanger to a i
low pressure consistent with No. 1
=
seal leakoff back-
'o pressure requirements.
Relief valve PSV-4121 I
i opens in seal return line to release
{
pressure to PRT.
l 9.
Air diaphragm a.
Falls closed.
a.
Charging and Volume a.
No automatic makeup of Control - seal water seal water to seal i
operated diaphragm valve flow.
standpipe that services No. 3 seal 1-LV-181, of RC pump No. 1.
No.
+
(1 -LV-180, offect on operation I
1-LV-119, and to bring the plant to
(
1-LV-178 are safe shutdown analogoas) conditions.
~.
90260:15/022285
APPENGIX A (Cont)
CVCS operation Effect on system
- pt
.Fa11ere Mode Function Goeratten and Shutdown Addittenal Information**
t
'b.
Falls open.
b.
Charging and Volume b.
Overf111 of seal water Control - seal water standpipe and dumping flow.
of reactor makeup water to containment sump during automatic makeup of water for No. 3 seal of RC pump No. 1.
No effect on operations to bring 4
reactor safe shutdown condition.
l i
i
- 10. Motor operated a.
Falls open.
- a. ' Charging and Volume a.
Fa11ere has no offact Control - seal water on CVCS operation globe valve flow and excess let-during normal plant NV-0112 (0100 down flow.
operation. However, f
analogous) under accident r
conditions requiring containment isolation of seal water flow and excess letdown flow, redundancy is reduced.
l i
b.
Falls closed.
b.
Charging and Volume b.
RC pump seal water re-Control - seal water turn flow and excess t
flow and excess let-letdown flow blocked.
down flow.
Fa11ere inhibits use of the excess letdown l
=
i fluid. system of the l'
l CVC5 as an alternate i
system that may be used for letdown flow l
l i
i i
90260:10/022285
4 AppfM0tX A (Cont) i CVCS operation Effect on Sy, stem oseenent Fallure Mode Function Operation and Shutdown Additional Information**
I control during normal plant operation.
Rellef valve pSV-8121 provides capability i;
of seal water to coollag RC pump bearings.
t
- 11. Motor rated a.
Falls open.
a.
Charging and Volume a.
Failure has no effect Control - charging on CVCS operation l
l gate va ve NV-8105 during normal plant flow.
(8106 analogous) operation. However, under accident l
condition requiring i
i 1 solation of charging l
[
Ilne, failure reduces redundancy of pro-viding isolation of
.[
normal charging flow.
3 f
b.
Falls closed.
h.
Charging and Volume b.
Failure prevents use Control - charging of normal charging flow.
Inne to RCS for
'boration, d11stion, and coolant makeup operations. Seal i
water injection path remains available for boration of RCS to a safe shutdown con-t centration level and j
makeup of coolant Juring operatlens to bring the reactor to i
conditions.
~
i 80260:10/022285 f
APPEN0lX A (Cont)
CVC5 operation Effect on System egeneet Failure Mode Function coeration and shutdown A441tlesal leforsetten**
I2.* Air diaphragm a.
Falls open.
a.
Charging and Volume a.
Failure prevents-
-Control - seal water.
manual adjustment at operated globe flow.
.C8 of charging flow, valve NCV-182 resulting in increased charging flow and decrease seal injection flow.
- b.. Fails closed.
h.
Charging and Volume b.
Failure prevents.
Control - charging normal charging flow.
flow.
Seal injection path remains for boration i
to safe shutdown.
- 13. Motor operated a.
Falls open.
a.
Charging and Volume a.
Failure has no effect i
Control - charging on CVCS operation globe valve NV-8110 flow and seal water during normal plant (tillA and 81118 flow.
operation. However.
analogous) under accident con-dition requiring j
i isolation of centrifugal charging pump mintflow line, fa11ere reduces re-dundancy of providing 1 solation of mintflow te section of pump via seal water heat exchanger.
j I
t l
I an7sa.in/022285 l
- ~, _ _ _. _ __
1.
j APPENetK A (Cont)
CVCS Operation Etfect on System component Fa11ere Mode Function Operation.and Shutdown Additional Infor1petion**
h.
Falls closed.
b.
Charging and Volume b.
Failure blocks mint-Control - charging flow to suction of flow and seal water centrifugal charging flow.
pumps via seal water heat exchanger.
j Normal charging flow and seal water flow i
prevents deadheading of pumps when used.
Boration of RCS to a safe shutdown concentration level and makeup of coolant during operations to bring reactcr to safe shutdown condition is 1
st111 possible.
i
- 14. Motor operated a.
Falls open.
-a.
Charging and Volisme a.
Failure has no offact globe valve MV-8146 Control - charging on CVCS operetten flow.
during normal plant
[
j e,
operatten, or safe a
shutdown operation.
valve is used during cold shutdown f
operatten to isolate normal charging line when using the l
aux 111ary spray l
during the cooldown I
of the pressurlzer.
i Cold shutdown of reactor is still l
l i
possible, however, time for cooling down PRZ will he extended. - -
I oneen. i n mmns I
~
=
1 APPENelX A (Cont)
CVCS Operation Effect on System Ceapopeat Failure Mode Function Goeration and Shutdown AdditionaUnfometton**
b.
Fails closed.
b.
Charging and Volume
'b.
Failure blocks nomal Control - charging charging flow to the flou.
RCS. Plant operator can maintain charging flow be establishing flow through j
alternate charging path by opening of isolation valve (NV-8147).
- 15. Noter operated a.
Falis closed.
a.
Charging and Volume a.
Failure reduces re-Control - charging dondancy of charging globe valve NV-8147 flow.
flow paths to RCS.
No offect on CVCS 4
operations during 7
7 nomal plant 3
operation, or safe shutdown operation.
Nomal charging flow path remains avall-able for charging flow.
i b.
Fa11s open.
b.
Charging and Volmine b.
Same offact on system Control - charging operation and shutdown
- flow, as that stated above for item No. 14, i
fa11ere mode ' Fails open" if alternate charging line is in l
use.
l 6
,o,,,..
.._..-_m.
APPENSIK A (Cont)
CVCS Operation Effect on System esponent Failure Mode Function Operation and Shutdown Additional lafetWatton**
- 16. Air diaphragn a.
Falls open.
a.
Charging and Volume a.
Failure results in Control - charging inadvertent operation operated globe flow.
of aux 111ery spray
. valve NV-8145 that results in a re-ductien of PRZ pressure durieri normal plant epera-tion. PRZ heaters ~
operate to maintsin required PRZ pressure. Soration of RCS to a safe shutdown concentratten level and makeup of coolant during operation to bring reacter to safe shutdown is still l
b possible.
~b.
Falls closed.
b.
Charging and Volume b.
Failure has no offect' a
Control - charging on tvCS operation fIow.
during normal plant
[
operation. Valve is used during cold i
shutdown operatten to r
activate auxillary spray for cooling t
down the pressurtzer after operation of a
RMS.
no?60:10/022285 t
.._ _. _. _ _......_.._ _. _.~. _.._.. _ --..._._.. _._ - -..-...._._ _ _._.
7..
AppENgIK A (Cont)
CVCS Operation Effect on System tempeammt Fallare nedt Functlen entratten and shutdeus Mditlemal lafersatten**
- 11. positive 31s--
a.
Falls to a.
Charging and volume a.
Failure reduces re--
deliver teatre) - charging dondency of providing placement Pump working fluid flew and seal water charging and seal water flew to RCS.
flow.
Centrifugal charging pumps provide al-ternate means of providing flow.
- 18. Centrifugal a.
Falls to a.
Charging and Volume a.
Failure reduces re-charging puup A deliver Centrol - charging dundancy of providing (pump g analogous) working fluid.
flew and seal water charging and seal
- flow, water flew to RCS. No l
effact on normal plant operatten, or bring reactor to a safe i
shutdeun conditlen.
Two centrifugal pumps are provided, both of which are backups to the normally running l
i t
positive displacement-pump.
t Ig. Air diaphragm a.
Falls closed.
a.
Cheetcal Centrol a.
Fa11ere blocks purifttation and hydrogen flew to VCT valve PCV-4156 Makeup - exygen resulting in less of l
operated globe control.
hydrogen ever-i pressure. No effect en operation to bring i
the reacter to safe shutdeun condition.
t l
nowntin/n306n5
ApptNSIX A (Cent) i CVCS Operetten Effect on Systes Failure Rada
" F = iten W ratlea and Shutdeum Addittenal lafernatten**
- g
- 20. Noter operated
.a.
Falls open.
a.
Charging and Volume-a.
Failure has no effect Centrol - charging en CVCS operatten gate valve LV-1128 flew sad seal water during neraal plant i
(LV-112C Ilaw.
operetten, and analogous).
bringing reacter to a safe shutdown conditten. Newever, under accident condittens requiring l
1solatten of VCT, i
fa11ere reduces re-dundancy of providing 1 solation for discharge line of VCT.
b.
Falls closed.
b.
Charging and Volume b.
Failure blocks fluid Centrol - charging flew from VCT during flew and seal water moraal plant opera-flow.
tien, and when bringing the reacter i
to safe shutdown conditten. Alternate-supply of berated (2000 ppa) coolant from the RMST to settlen of charging pumps can be es-tablished from the centrol room by the operater through the epening of RWST l
1seletten valves (LV-lits and LV-112E) and starting a centrifugal ch rging pump.
._ ~..-
1 APPENDIR A (Cont)
I cvCS operation Effect on System omeonent Failure Mode Function Operation and Shutdown Additional Information**
'1. Air diaphragm
'a.
Falls closed.
a.
Chemical Control, a.
Failure blocks vent-Purification and ing of VCT gas stature operated valve
-)
PV-115 '
Makeup - oxygen to gas waste pro-control.
cessing system for stripping of fission products from RCS coolant during normal plant operation. No efiect on operations to bring the reactor to safe shutdown condition.
i t
- 22. Air diaphragm a.
Falls closed.
a.
Boron Concentration a.
Failure blocks fluid Control - reactor flow from reactor l
operted i
diaphragm valve makeup control -
makeup control system boretton, auto for automatic boric FV-110s makeup, and alternate acid addition and I
d11stion.
reactor water makeup during nors;al plant operation. F311cre also reduces redundancy of fluid i
flow paths for dilution of the RCS coolant by reactor makeup water and blocks fluid flow for boration of the i
reactor coolant when j
bringing the reactor i
to a safe shutdown j
condition. Boration (at SA storage tank I
i I
l anim. I n /n??7n%
~
1 APPENOIR A (ront)
~
.CVCS Operation Effect on System Component Fa11ere Mode Function
[beration and Shutdown Additional Information**
boron concentration level)~of RCS coolant.
is possible by opening of alternate SA tank isolation valve (nV-s104) at Co.
i b.
Falls open.
b.
Borm Concentrat W b.
Failure allows for l
Control - reactor alternate dilute makeup control -
mode type operation
[
boration, auto for system operation makeup, and alter-of normal dilution of nate dilution.
RCS coolant. No j
effect on CVCS eperation during normal. plant operation i
and bringing the reactor to a safe I
shutdown condition.
- 23. Air diaphragm ~
a.
Jalls closed.
a.
Doron Concentration a.
Fa11ere blocks fittfd i
Control - reactor flow from RHCS for operated diaphragm valve makeup control -
dilution of RCS cool-d11ution and alter-
' ant during normal FV-ille nate dilution.
plant operation. No l
etfect on CVCS I
operation. Operator can d11ste RCS coolant
[
by estab-11shing " alternate dilute
- node of system t
operation.
i i
8026Q:10/022285
APpEN0!K A (Cont)
CVCS Operatlon Effect on System Operation and Shutdown Additional Information**
Coseenent Fallure Mode Function b.
Falls open.
b.
Boron Concentration b.
Failure allows for Control - reactor alternate dilute I
makeup control -
mode type operation dilution and alter-for system operation nate dilution.
of boration and auto makeup of RCS coolant. No offact on l
CVCS operation during normal plant operation or shutdown.
- 24. Air diaphragm a.
Falls open.
a.
Soron Concentration a.
Failure prevents the 1.
Valve is operated globe Control - reactor addition to a pre-designed to valve FV-Il0A makeup control -
selected quantity of fall open.
boration and auto concentrated boric makeup.
acid solution at a 2.
Valve position preselected flow rate indication (open to the RCS coolant to closed 7
during normal plant position change)
G operation. End when in control room; bringing the reactor and boric acid to a safe shutdown flow recording condition. Deration*
(fR-110) and to bring the reactor flow deviation to a safe shutdown alara at the condition is possible, control board, however, flow rate of solution from the BA storage tank cannot be automatically controlled.
80260: 10/022285
APPEN9tX A (Cont)
~
CVCS operation
.Effect on System Coseopent Failure flode Function eseration and Shutdown Additional Information**
b.
Falls closed.
b.
toren Concentration b.
Failure blocks fluid 1.
BA flow Control - reactor flow of boric acid deviation makeup control -
solution from BA stor-alarm and boration, and auto age tanks during makrup.
total makeup i
makeup.
Beration (at BA stor-flow devi-l age tank boren concen-atton alarm.
l tration level) of RCS coolant is possible by opening of alternate SA tank 15:lation valve i
(MV-5104) from control room.
~
- 25. Air diaphragm
-a.
Falls closed.
a.
toren Concentration a.
Failure blocks fluid i
Control - reactor flow of water from operated globe makeup control -
reactor makeup valve FV-Illa d11ste, alternate control system during dilute and auto normal plant oper-
>4
- makeup, atton. No cffect on system operation.
b.
Falls open.
h.
Boron Concentration b.
Failure prevents tho' l-l Control - reactor addition of a pre-makeup control -
selected quantity d11ste, alternate of water makeup at a i
dilute and auto preselected flow makeup.
rate to the RCS coolant during l
normal plant operation.
l 90260:10/022285
__ ~...._..__.._..
APPEN0tX A-(Cont) f CVCS Operation Effect on System Failure nada F---- ttaa Gperaties and ShutdeWe Addittemal Infomatten**
l tampament
- 26. Noter operated a.
Falls closed.
a.
toren Concentration a.
Fallure reduces re-1.
Valve is at a Control - reactor dundance of flow paths closed position l
globe valve NV-0104.
makeup centrol -
for s og boric during normal i
beratten and aute acid so en from BA RNCS operatten.
store e tank to RCS via l
makeup.
2.. Valve pestilen chart ne pumps.
j I
Normal flew path via indication RMCS remain available (closed to open for beratten of RCS positten change) ion and flew indicat
- coolant, (FI-183A) at con-trol board.
l
~
3.
If both flew paths from the beric acid storage tank are s
blocked due to fallure of t
T 14elatten valves l
(FV-110A and NV-8104), berate (2000 ppm) from i
RWST is available opening 1selatten valve LV-1128 er LV-112E.
b.
Fa11s open.
b.
toren Concentratten b.
Fallure prevents the Control - reactor additten of a pre-I l
makeup control -
selected quantity of i
beretten and auto concentrated beric j
acid seletion at a
- makeup, preselected flow rate to the RCS coolant.
8eratten is possible, i
an nn in/n30605 l
-.. -. - - -. ~. -....
l APPENGtX A (Cont) t CVCS Operatten Effect en System Campensat Failure nada Functlea Geeraties and shutem a Additional Info ina11se**
j however, the flew rate l
from the beric acid storage tank cannet be automatically controlled.
i I
- 27. Seric acid a.
Falls to a.
toren Concentratten a.
Failure prevents re-transfer pump deliver Centrol - reacter dundancy of delivering Pump 1206-P6-006 we-1 ting fluid.
~ makeup control -
15 a solutten to CVCS.
beratten and ante Alternate BA transfer l
(Pump 1200-P6-007 makeup.
pump may be used i
stellar) to provide necessary delivery of wetting fluid for CVCS system l
operatten.
i l
- 29. Air diaphragm a.
Falls open for a.
Charging and Volume a.
Fa11ere bypasses neraal
. operated three-flew only to Centrol - letdown down flew SRS recycle way valve LV-112A 8R5 recycle flow.
heldup tank resulting In eacessive use of RMCS.
l heidup tank.
I 2g. Air diaphragm a.
Falls open.
- a. Charging and Volume a.
Fallure prevents manual 1.
Valve is designed j
Centrol - charging adjustment at CS of to fall open.
operated globe flow and seal water charging ficu from the valve FV-121 flow.
centrifagal charging pumps. Speed centrol of positive displacement l
pump is not affacted.
I i
l l
t
l l
APPENSIK A (Cont) l.
b
(
CVCS Operation Effect on System Component Failure Mode Function Operation and Shutdown Additional Information**
b.
Falls closed.
b.
Charging and Volume b.
Failure blocks normal Control - charging path for use of~
l.
flow and seal water centrifugal charging flow.
pumps for charging and seal water flow.
Manual valve 151 or-152 can be open to allow flow from the corresponding pump. A path from the positive displacement pump would also still be available.
- 30. Manual Valve a.
Falls open.
a.
Charging and volisme a.
Fa11ere provides un-Emergency beration control-emergency borated water flow flow indication t
183 boration 11ne flush to centrifugal (FI-183), VCT from Reactor Water charging pump A level indication askeup suction.
(LI-L12) and high level A,
alans at Ca.
l w
[
- 31. Dominera11rer a.
Falls open.
a.
Moraal plant Opera-a.
Fallure during restn*
lVCT level Indication tion-open. Resin flush with reactor (LI-112) and high Flush Operation-
' makeup water would level alare at CB.
i Olscharge Valves 65 Closed.
provide unborated i'
(63,74, water to charging 259, 260, 26) pump section.
262, 263 analogous) t h
l i
8026g:10/022285 i
l APPENetX A (Cont)
CVCS Operatten Etfett en System Egsmonant Fallere Mode Function Operatten and Shutdown Additionel Inferentles**
l
- 32. Manual Valves 116 a.
Operator falls a.
Chemical addition a.
Operater error VCT level indicatten i
and 181.
to close.
make up water supply (fallure to secure (t.1-112) and high i
and discharge valves.
chemical additten) level alara at ca.
provides unberated Flow orifice restricts l
water to charging dilutten flew to 3.5 gpe i
(
pump section.
er less.
f 33, 31aphrage Valve a.
Falls closed.
a.
Seric Acid supply to a.
Closure during make-SA flew deviatten transfer pumps up terminates borated alarm and total makeup during makeup.
water flew to CVCS.
flow deviatten alare.
NV-8461.
l t
- 34. Manual Valve 226 a.
left open a.
Dominera11rer a.
Reacter makeup water i
flushing would be admitted inte i
BTRS lines l
l s
l y
POTES:
I
- l.lst of acronyms and abbreviattens used:
84
- Soric Acid BR5
- Seren Recycle System BTR
- Seron Thermal Regeneration j
j 01R5
- Seren Therest Regeneratten System Cs
- Centrel seard CNS
- Chemical and Volume Control System i
Domin.
- Seminere11rer l
- Heat Euchanger l
PRZ
- Pressertzer RC
- Reacter Coelant RCS
- Reactor Coelant System i
- Residual Neat Removal System I
RMST
- Refueltag Water Storage Tank RMCS
- Reacter Makeup Centrol System VCT
- Volume Control Tank
- Baron Recycle System Recycle Holdup Tank RHT Additional Information supplied when failure is significant in boron dlistion analysts.
~ ~ ~ " ' " ' " ' " " " "
APPENDIX 8 I*
RESPONSE TIME CALCULATIONS 1.
L This appendix presents the derivations of the equations and the parameters used in the calculations of response times.
1 i
j I.
Time to Criticality (T )
g 4
A differential equation representing t6e rete of change of boron concentration is a mixing volume can be written as:
i l
1 6
j where: V
= RCS volume (gallons) l C
= boron concentration in the RCS (ppa) f C, = M n concentration of injec W m er (p W 4
j Q
= dilution flow rate (gallons / minute)
Assuming that the injected water contains no boron (C,= 0), Eq. 1 g
j becomes:
1 1
dc/dt = -Q/V C Eq. 2 i
Integrating gives the following:
i Eq. 3 l
C = C,exp (-Q/V t) l Where C, is the initial boron concentration. At time t.
g i
f C, = C,exp (-Q/V t )
Eq. 4 g
Where C, is the critical boron concentration. Solving for t,
e t = V/Q In (C,/c )
)
g g
b 80260:10/061885 8-1 j
1 i
II. Time to High Flux at Shutdown Alarm The time to this alars is the time required to increase the neutron count to 3.16 times background. The neutron flux, and thus the count rate, is inversely proportional to K,ff.
If 5,is the neutron flux at t = o and K,ff(o) is the initial K,,, value.
Eq. 6 5/0, = (1-K,,,(o))/(1-K,f f)
The following equation can also be written:
Eq. 7 K,ff = 1 - (C-C,)B where B is boron worth. This is assuming that 8 is constant with respect to boron concentration, which is valid for relatively small changes in concentration (a few hundred ppm).
i From Equations 6 and 7:
Eq. B S/5,= ( C,-C,) / ( C - C,)
Substituting Eq. 3 into Eq. 3:
Eq. g 5/0, = ( C,-C,) / [ C, exp (-Qt/V) - C,)
l Solving for the time of the high flux alarm (S/5,= (10)
T(HF) = - V/0 in [.3162 (1-C,/C,) + C,/C, )
i i
l f
f B-2 S0260:10/062085
III. Time to High VCT Level Alam The VCT high level alarm sounds when 2167.1 gallons are in the VCT. Assuming l
that the VCT is. operating at the low level setpoint an increase of 1366 l
gallons is required to obtain the alare. This is conservative, since the vcT With a l
initial level would normally be well above the low level setpoint.
dilution flowrote of Q gps, the time to the high level alare is:
I i
t = 1366 / 0 f
IV. Soron Dilution Parameters l
A detailed review of the CVCS system and associated procedures was performed j
and potential boron dilution events identified. For each of these initiators j.
'the resulting dilution flow rate was calculated, and it was conservatively j
j assumed that this entire dilution flowrote was supplied to the RCS.
l l
Additionally conservative estimates of the effective volume of the RCS were l
calculated (e.g. - In modes 3 or 4 with one RCP running, no credit was taken l
for backflow through the idle RCS loops). The cycle 1 core characteristics of f
the core were' reviewed.and the limiting conditions of critical boron These calculations concentration and boron worth obtained for each mode.
l conservatively assumed beginning of life Xenon free conditions with the most l
reactive control rod fully withdrawn. Vogtle Electric Generating Unit has strict administrative controls to preclude the possibility of initiators 1, 2 l
and 4 in mode 5 when the pressurizer level is below the bottom of the Therefore mode 58 (RCS volume of 3435 ft ) was analyzed 3
f indicated range.
only for initiator number 3 (chemical addition). The parameters used in the boron dilution analysis are summarized in Table 5-1.
I V.
Results i-The
[
Table B-1 lists the parameters used to calculate the response times.
critical boron concentrations were obtained from core physics calculations l
performed for a very similar reactor core. The critical boron concentrations used are fo_ beginning of life Xenon free conditions and assume that the most r
i reactive rod is fully withdratm.,
i i
80260:10/061885 8-3 b
v
s The table Table 8-2 shows the results of the response time calculations.
includes the time to alam annunciation of the alarm which first warns of the l
boron dilution event. Greater than fifteen minutes is available between boron dilution annunciation and criticality for all cases.
j I
ap T
1 r
l l
1 t-
[
C-i
.80260:10/061885 B-4 h..
TA8LE 8-2 RESPONSE TIME CALCULATIONS
{
g T,(A)2 3
T
-T INITIATOR FLOWRATE MODE yp g
1 63 3, 4 U2 87 23(VL) 45 j
2 120 3, 4 69 46-12(VL) 23 3
3.5 3, 4 2368 1570 1570(HF) 798 4
186 3, 4 44.5 29.5
<1(FD) 15 1
63 5A 138 93 23(VL) 45 2
120 5A 73 49 12(VL) 24 3
3.5 5A 2487 1672 1672(HF) 815 4
186 5A 46.8 31.5
<1(FO) 15.3 3
3.5 58 1463 983.
983(HF) 480 t
Notes:
l-(1) Tc is the time to criticality in minutes l
(2) Ta is the time to the annunciation of an alarm which has appropriate procedures to terminate the dilution event.
'A' represents the alarm which is applicable for the specific dilution initiator:
F0 = Soric Acid and Makeup Flow Deviation Alarms VL = VCT high level alarm HF = High Flux at Shutdown Alarm i
(3) Bilution Flowrates into the CVCS system in gps.
i i.
il' 80260:10/062005 8-5
P TABLE 8-1 PARAMETERS 811ution Flowrates:
(Source - References 1 and 2)
Initiator.
Flowratefans) 1 1
63 2
120 3
3.5 4
186 Volumes:
(Source - Reference 1)
- gggg, Volume (ft J, Volume (aal)
- 3. 4 5840.0 43800 1
Sa (filled) 11200.9 84007 Sb (drained) 3435.0 25163 Soron Concentrations:
Mode C (ppm)
C,(ppe) 5 SOM c
3, 4 '
720 870 2.0 5
971 1077 1.5 j
1 80260:10/062005 8-6
t APPEN0!X C i
i OATA i
l This appendix presents the component failure rates which were used in this study. Component failure rates are addressed first, followed by example calculations of alarm reliabilities.
P - anent Failure Rates Conoonent Failure Rate jon
\\
I j
Level or Flow detectors 4.7 E-6 / hour NUREG/CR-2771 j
Air operated FCV fails open 1.5 E-6 / hour NURES/CR-2770 I
Source range flux detector 5.0 E-6 / hour NURES/CR-2771 l
Instrument signal conditioning system 7.5 E-6 / hour NURES/CR-2771 i
Alam bistable 8.2 E-7 / hour IEEE-500 Alam annunciator fails to operate 1.0 E-6 / hour IEEE-500 Air operated valve fails to close 2.0 E-3 / demand NUREE/CR-1363 Notes:
(1) Includes both inoperability faults and reduced capability faults Example alam M11 ability calculations:
High Flux Alarm Reliability:
Failed detector:
5.0E-6 x 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> = 3.0 E-05 (1)
Failed signal conditioning system: 7.5E-6 x 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> = 4.7E-05 Alarm bistable failure: 8.2E-7 x 15 x 24 = 3.0E-04 (2)
Alarm failure: 1.0E-6 x 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> = 1.2E-05 (3)
Total HFA unavailability
= 3.gE-04 90260:10/061R95 C-I
?
Notes:
}
(1) It was assumed that any detector or signal conditioning fault would be detected within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> either during the channel check or by l
comparison with the redundant detector indications.
No additional credit was taken for the redundant channel.
(2) Analog channel operational test performed monthly.
f (3) Alars annunciator tested daily.
1 Volume Control Tank Level Alars Reliability; j
1 Failed detector: 4.7E-6 x 360 = 1.7 E-03 (1) f Failed signal conditioning system: 7.5E-6 x 360 = 2.7E-03 Maintenance unavailability: 1.2E-5 x 8760 x 100 / 8760 = 1.2E-3 Alarm bistable failure: 8.2E-7 x 100 x 24 = 3.5E-03 l
l Total VCT 1evel alars unavailability: g.1 E-03 l
Notes:
i l
(1) Any detector or signal conditioning failure was assumed to be detected within 15 days,of the failure either by test or by comparison with the redundant. detector.
l (2) VCT level alars histable assumed tested annually.
(3) Assumed a mean time to repair of 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br />.
.80260:10/061885 C-2
i UselTED STATES f.,
/y mess,q NUCLEAR REoULATORY COMMISSION J
p' Raoion si.
e 101 MARitTTA STREET, N.W.
ATLANTA, GEORGI A 30323 DEC 17 W l
Docket No. 50-424-U.c$
License No. NPF-68 T'
l55 g
Georgia Power Company ATTN: Mr. James P. O'Reilly Senior Vice President-Nuclear Operations t
i P. O. Box 4545 Atlanta, GA 30302 1
Gentlemen:
i i
SUBJECT:
INSPECTION REPORT NO.'50-424/87-60 This refers to the Nuclear Regulatory Commission (NRC) inspection conducted by Messrs. J. F. Rogge, C. W. Burger, and R. J. Schepens on October 8 -
November 20, 1987.
The inspection included a review of activities authorized for your Vogtle facility. _ At the conclusion of the inspection, the findings 1'
were discussed with those members of your staff identified in the enclosed L
inspection repcrt.
t Areas examined during the inspection are identified in the report.
Within these areas, the inspection consisted of selective examinations of procedures l
and representative records, interviews with personnel, and observation of activities in progress.
Within the secpe of the inspection, no violations or deviations were identified.
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In accordance with Section 2.790 of the NRC's " Rules of Practice," Part 2, l
Title 10, Code of Federal Regulations, a copy of this letter and the enclosure will be placed in the NRC Public Document Room.
Should you have any questions concerning this letter, please contact us.
Sincerely.
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Virgilvl. Brownlee, Chief Reactor Projects Branch 3 Division of Reactor Projects
Enclosure:
?NRC. Inspection Report
'cc w/ enc 1:
(See page 2)
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DEC 171967 Georgia Power Company cc w/ encl:
P. D. Rice, Vice President, Project Director C. W. Hayes, Vogtle Quality Assurance Manager G. Bockhold, Jr... General Manager, Nuclear Operations L. Gucwa, Manager, Nuclear Safety and Licensing J. A. Bailey, Project Licensing Manager B. W. Churchill, Esq., Shaw, Pittman, Potts and Trowbridge D. Kirkland..III, Counsel, Office of the Consumer's Utility Council D. Feig, Georgians Against Nuclear Energy l
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Report No.:
50-424/87-60 Licensee: Georgia Power Company P. O. Box 4545 Atlanta, GA 30302 Docket No.: _50-424 License No.: NpF-68 Facility Namei Vogtle 1 Inspection Conducted:
October 8 - November 20, 1987 i
n Inspectors: N Uw,A IbaNo-ac,4
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,p4.5Rogge,SeniorResidentInspector Date Signed k.G, 8 f.N ca.
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I ? / /G l9 7 p R. J. Schepens, Resident Inspector Date Signed
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8 & A 8nn A i2//s(P7 P LC. W. Burger, Resiaant Inspector Date Signed i ({
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. Approved by:
M. 'V.'
Sinkule, Section Chief Date' Signed Division of Reactor Projects
SUMMARY
Scope:
This routine, unannounced inspection entailed resident inspection in the following areas: plant operations, radiological controls maintenance, surveillance, fire protection,
- security, and quality programs and administrative controls affecting Quality.
Results:
No violations or deviations were identified.
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REPORT DETAILS 1.
Persons Contacted Licensee Employees G. Bockhold, Jr., General Manager Nuclear Operations
- T. V. Greene, Plant Support Manager
- R. M. Bellamy, Plant Manager "E. M. Dannemiller, Technical Assistant to General Manager C. C. Echert, Technical Assistant to Plant Manager "J. E. Swartzwelder, Nuclear. Safety & Compliance Manager
- W. F. Kitchens, Manager Operations
-l R. E. Lide, Engineering Support Supervisor "H. Varnadoe, Plant Engineering Supervisor
- R. E. Spinnatu, ISEG Supervisor C. W. Hayes, Vogtle Quality Assurance Manager
- G. R. Frederick, Quality Assurance Site Manager - Operations W. E. Mundy, Quality Assurance Audit Supervisor M. A. Griffis, Maintenance Superintendent
- R. M. Odom, Plant Engineering Supervisor
- C. L. Cross, Senior Regulatory Specialist S. F. Goff, Regulatory Specialist l
- A. L. Mosbaugh, Assistant Plant Support Manager H. M. Handfinger, Assistant Plant Support Manager F. R. Timmons, Nuclear Security Manager Other licensee employees contacted included craftsmen, technicians.
supervision, engineers, operations, maintenance, chemistry, inspectors, and office personnel.
- Attended Exit Interview 2.
JExit Interviews (30703)
The inspection scope and findings were summarized on November 20, 1987 i
with those persons indicated in paragraph 1 above.
The inspector described the areas inspected and discussed in detail the inspection results. _ No dissenting comments were received from the licensee.
The licensee did not identify as proprietary any of the materials provided to or reviewed by the inspector during this inspection.
Region based NRC exit interviews were attended during the-inspection period by a resident inspector.
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4 3.
Operational Safety Verification (71707)(93702) l The plant began this inspection period in Power Operation (Mode 1) at 100%
power until October 9 when the unit was tripped to complete a portion of j
the startup testing program and commence a short outage.
The outage proceeded without difficulty until the number 1 reactor coolant pump motor failed. As a result of the failed motor, Unit I restart was delayed approximately seven days.
The unit entered Hot-Standby (Mode 3) on 1
October 27.
Shortly after achieving Mode 3 the residual heat removal j
crosstie. valve motor operator failed and the engineering walkdowns i
identified that the reactor vessel level instrument impingement plates 4
were not installed.
These two problems resulted in further startup delays.
On October 31 the unit entered Startup (Mode 2) and achieved Mode 1 on November 1.
The unit achieved 100% power on November 4.
On
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November 5 the unit tripped on a turbine trip when a vibration sensor was bumped.
The unit returned to Mode 1 on November 6 and achieved 100% on Novehiber 7.
On November 9 the unit performed the 10% load swing startup i
test. On November 11 the unit tripped from 100% reactor power when the l
wrong test panel was used during the performance of a reactor trip breaker test. On November 12 the unit returned to Mode 1 and achieved 90% power.
F om November 12 through 17 the unit experienced secondary water chemistry problems which limited power and required the plugging of condenser tubes.
i On November 18 the unit was held at 98% power while engineering concerns in regard to exceeding the 3411 MWT limit 'were resolved. On November 19 the unit achieved 100% power. The plant experienced three ESF actuations; the Control Room Emergency Ventilation System on October 26 when a l
s technician improperly reset the radiation monitors and on November 17 when RE-12116 spiked high, an auxiliary feedwater actuation on November 5 when i
l an operator shut the discharge valve of the running condensate pump due to improper labeling, and a Containment Ventilation Isolation from RE-2565 on November 9 when the check source did not fully retract.
A Notice of l
Unusual Event was reported on November 17 when power was lost to meteorological instruments.
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Control Room Activities l
Control Room tours and observations were performed to verify that j
facility operations were being safely conducted within regulatory requirements.
These inspections consisted of one or more of the following attributes as appropriate at the time of the inspection.
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i Proper Control Room staffing Control Room access and operator behavior Adherence to approved procedures for activities in progress 4
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i Adherence to Technical Specification (TS) Limiting Conditions for Operations (LCO)
Observance of instruments and recorder traces of safety related and important to safety systems for abnormalities Review of annunciators alarmed and action in progress to correct I
Control Board walkdowns Safety parameter display and the plant safety monitoring system i
operability status Discussions and interviews with the On-Shift Operations Supervisor, Shift Supervisor, Reactor Operators, and the Shif t i
Technical Advisor to determine the plant status, plans and i
assess operator knowledge l_
Review of the operator logs, unit log and shift turnover sheets l
No violations or deviations were identified.
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Facility Activities i
Facility tours and observations were performed to assess the effectiveness of the administrative controls established by direct observation of plant activities, interviews and discussions with licensee personnel, independent verification of safety systems status and LCO's, licensee meetings and facility records.
During these inspections the following t'bjectives were achieved:
(1) Safety System Status-(71710)
Confirmation of system 1
operability was obtained by verification that flowpath valve alignment, control and power supply alignments, component conditions, and support systems for the accessible portions of j
l' the ESF trains were proper.
The inaccessible portions are i
confirmed as availability permits.
Additional in-depth inspection of the Auxiliary Feedwater System was performed to review the system lineup procedure with the plant drawings and as-built configurations, compare' valve remote and local indications, and walkdown of hangers, supports, snubbers and electrical equipment interiors. The inspector verified that the lineup was in accordance with license requirements for system j
operability.
l (2) Plant Housekeeping Conditions Storage of material and components and cleanliness conditions of various areas 4
throughout the facility were observed to determine whether safety and/or fire hazards existed.
(3)
Fire Protection Fire protection activities, staffing and L
equipment were observed to verify that fire brigade staffing was appropriate-and that fire alarms, extinguishing equipment, actuating
- controls, fire fighting equipment, emergency i
equipment, and fire barriers were operable.
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(4) Radiation Protection (71709) - Radiation protection activities, staffing and equipment were observed to. verify proper program j
implementation.
The inspection included review of the plant program effectiveness.
Radiation work permits and personnel 4
compliance were reviewed during the daily plant tours.
Radiation Control Areas (RCAs) were observed to verify proper j
identification and implementation, a
(5) Security (71881) - Security controls were observed to verify that security barriers were intact, guard forces were on duty, and access to the Protected Area (PA) was controlled in 1
accordance with the facility security plan.
Personnel within the PA were observed to verify proper display of badges and that l
personnel requiring escort were properly escorted.
Personnel j
within vital areas were observed to ensure proper authorization for the area. Equipment operability and proper compensatory 2
l activities were verified on a periodic basis.
t (6) Surveillance (61726)(61700) - Surveillance tests were observed l
to verify that approved procedures were being used; qualified personnel were conducting the tests; tests were adequate to verify equipment operability; calibrated equipment was utilized; and TS requirements were followed.
The inspectors observed, portions of the following surveillances and reviewed completed j-data against acceptance criteria:
Date Surv. No.
Dept.
Title 11/3/87 14915-1 Ops QPTR Special Condition Surveillance Log 11/4/87 14915-1 Ops Control Rod Insertion Limits Special Condition Surv. Log 11/4/87 14205-1 Ops Plant Emergency Signal Weekly Operability Test 11/4/87 14805-101 Ops Quarterly, Train B RHR Pump &
Check Valve Inservice Test 11/6/87 14808-102 Ops Quarterly, Train B CCP &
Check Valve Inservice Test 11/19/87 14030-1 Ops Power Range Calorimetric Channel Calibration 11/20/87 14825-108 Ops Quarterly, Train A AFW Valve Inservice Test (7) Maintenance Activities (62703)
The inspector observed a
maintenance activities to verify that correct equipment clearances were in effect; work requests and fire prevention work permits, - as required, were issued and being fc11 owed; quality control-personnel were available for inspection activities as required; retesting and return of systems to 4
5 service was prompt and correct; TS requirements were being followed. The maintenance backlog was reviewed and noted as consisting of approximately 2.100 MWO's (i.e., both corrective and preventive) prior to the outage. Maintenance had scheduled 249 maintenance work orders to be worked during the outage.
During the outage the inspector observed that maintenance had actually performed an additional 151 MWO's due to discovery items and 109 MWO's due to the forced outage on the reactor coolant pump motor in addition to - the 249 MWO's planned for a total of 509 MWO's. At the completion of the outage the outage backlog had - been reduced from 506 to 300 MWO's, however the total 'MWO backlog had increased slightly from 2.100 to 2,178 MWO's.
The inspector either observed maintenance activities or reviewed completed maintenance work packages for the following l
maintenance activities:
MWO No.
Dept.
Work Description 1-87-02793 Elect. Maint.
Perform MOVATS Procedure and DCP VIE 007 1-87-05326 Elect. Maint.
Investigate Problem With Open Indication Light Not Working 1-87-08736 Mech. Maint.
Implement Design Change Package To Pressurizer Level Transmitter LT-461 i
1-87-11815 Maint./ Chem.
Condenser Waterbox B West Tube Leak Check & Plugging i
(8) Outage Activittes (71711)' - The inspector observed portions of the outage activities to determine management effectiveness in i
L conducting outages. While this was not a refueling outage it 4
did demonstrate the licensee's ability to schedule, prepare, and i
l execute the plan.
As noted above, at the completion of the j
outage the outage backlog had been reduced from. 506 to 300 MWO's. During the course of the outage teamwork was evident in i
surfacing new problems and achieving resolution to prevent a new critical path from developing.
The planned critical path wort i.tvolving the removal of the temporary steam strainers was l
achieved ahead of schedule. The outage work inside containment was performed with few difficulties. Two major items did occur which had severe schedule impact and resulted in a seven day
. restart delay.
These items were the motor replacement on the number one reactor ecolant pump and the failed motor on the RHR crosstie valve HV-87168.
Teamwork in resolving both problems resulted in a very ccordinated repair effort. Unit recovery was delayed upon - di scovery tnat the impingement plates for RVLIS were not installec~nor locatable, which required new pieces to I
be fabricated.
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6 No violations or deviations were identified.
4.
Review of Licensee Reports (90712)(90713)(92700) a.
In-Office Review of Periodic and Special Reports This inspection consists of reviewing the below Tisted reports to determine whether the information reported by the licensee is technically adequate and consistent with the inspector knowledge of the material contained within the report.
Selected material within the report is questioned randomly to verify accuracy to provide a reasonable assurance that other NRC personnel have an approp: tate document for their activities.
Monthly Operating Reports - The report dated October 8,1987 was reviewed. The inspector had no significant comments regarding these reports.
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Licensee Event Reports (LER's) and Deficiency Cards (DC's)
Licensee Event Reports (LER's) and Deficiency Cards (DC's) were reviewed for potential generic impact, to detect trends, and to determine whether corrective actions appeared appropriate.
Events which were reported pursuant to 10 CFR 50.72, were reviewed as they occurred to determine if the technical specifications and other regulatory requirements were satisfied.
In-office review of LER's may result in further followup to verify that the stated corrective actions have been completed, or to identify violations in addition to
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Each LER is reviewed for enforcement action in accordance with 10 CFR Part 2, Appendix C.
Review of DC's i
was performed to maintain a realtime status of deficiencies.
4 determine regulatory compliance, follow the licensee corrective r
actions, and assist as a basis for closure of the LER when reviewed.
Due to the numerous DC's processed only those DC's which result in enforcement action or further inspector followup with the licensee at the end of the inspection are discussed as listed below. The LER's j
denoted with an asterisk indicates that reactive inspection occurred at the time of the event prior to receipt of the written report.
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(1) Deficiency Card reviews:
DC 1-87-2616 "D5-416 Reactor Trip Breaker Inspections" This deficiency documents the results of the weld inspections.
i During the inspections the NRC resident and vendor branch inspectors were present.
The results of the inspection were acceptable however the NRC recommended that the shaf ts be replaced in the long term. These inspections were performed to address.the concerns as addressed in Information Notice No. 87-35.
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7 DC 1-87-2708 "RVLIS Impingement Covers" On 10/23/87 the impingement cover plates for RVLIS tubing for 1-LX-1310 and 1-LX-1320 were not installed. In order to correct this problem new plates were f'abricated and installed.
This resulted in a delay in return to. power.
DC 1-87-2733 " Control Room Isolation While Resetting Radiation Monitors" This DC describes an unplanned actuation on 10/26/87 when the radiation parameter resetting procedure did not call for blocking of the output.
In addition poor communication between operators and the chemistry department was exhibited in that the status of the Control Room Ventilation being reset was not fully understood nor was the nature of work to be performed.
DC 1-87-2753,1-87-2766,1-87-2846 " Mode 3 Entry Performed without all requirements met" These deficiency cards documented three instances that the licensee identified after the unit entered Mode 3.
The three cases were failure to perform IST testing on the A train AFW discharge check valve following maintenance, failure to have the Steam Driven AFW pump steam admission valves open, and failure to perform a functional test of the A train safety injection pump following changeout of the lubricant.
Each instances had minimal impact as follows:
the check valve tested satisfactorily, full secondary steam pressure had not been obtained to support the surveillance testing, and the safety injection pump was tested satisfactorily.
DC 1-87-2915 " Reactor. Trip While Performing OSP 14701-1" This Reactor Trip resulted when the B train auto shunt trip test panel was used during the testing of the A train breaker. While the procedure directed the operator to the correct test panel no labeling was in place at the test panel to indicate that the wrong train was being utilized. During the performance of the undervoltage coil trip test ta additional indication existed to indicate that the shunt coit nad not been blocked. When the shunt coil trip test was executed the B train shunt coil energized and the B train reactor trip breaker opened.
Since the A train SSPS was in test to support A train reactor trip breaker testing the control room operator had to insert a manual trip to open the A train reactor trip breaker and perform a manual. start of the A train Auxiliary Feedwater Pump.
DC 1-87-2974 " Missed Surveillance" This deficiency occurred on i
November 16 when a ' room temperature surveillance was not performed due-to ' the floor being painted. The operator NA'd the step which.was later identified during a supervisor review and at that time it was noticed that the TS had been missed.
(2) The following LER's were reviewed and are ready for closure pending verification that the licensee's stated corrective actions have been completed.
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(a) 50-424/87-05 Rev 0-4 "120V AC Voltage Transient Causes ESF Actuations" These LERs describe a plant condition where a voltage transient causes ESF actuations uoon energization of the Safety System Sequencer Panel.
The inspector noted to the licensee that the final supplemental LER was due on July 30. 1987.
The licensee informed the inspector that the LER will be closed on January 1988 once the information is received from Westinghouse.
(b) 50-424/87-20, Rev 0 "ESF Actuation Caused by Excessive Leakage Through a Main Feedwater Regulating Valve".The inspector noted to the licensee that the final supplemental LER was due on July 10, 1987.
The licensee informed the inspector that the LER will be closed once the final
. corrective action. is performed.
The LER states that further testing of_ valve 1HV-5139 will be performed when the unit is in Mode 3.
The Licensee failed to accomplish this test during the outage but will do the test at the next forced outage or refueling.
The final LER will be issued following the test.
(c) 50-424/87-56, Rev 0 " Technical Specification Not Met Due To Incomplete Vendor Software For Dose Calculations" This LER describes an event which occurred on September 16. 1987 when it was identified that the cumulative dose calculation program for gaseous releases to the atmosphere for radiciodines did not include isotope I-133 in the sof tware package. The licensee identified this during a data review while preparing the semi-annual radioactive effluent' release report.
Corrective action includes revising the software and the performance of a functional testing.
The inspector has no further questions regarding this report.
The following is identified:
50-424/ LIV 87-60-01 " Failure To Implement an Ap~propriate Surveillance to determine cumulative dose contributions in accordance with the ODCM per TS 4.11.2.3 - LER 87-56" (d) 50-424/87-58, Rev 0 " False Signal From Rad Monitor Leads To Control Room Isolation" This LER describes an event which occurred on September 21, 1987 when the control room isolation occurred due to a falso high radiation signal from 1-RE-12116. While no violations resulted from this event the licensee has.yet to specify the root cause of the I
failure in a supplemental report due December 15, 1987.
I (3)..The following LER's were reviewed and are considered closed.
(a) 50-424/87-01, Rev 0 " Incorrect Transmitter Circuit Board Leads to Missing a Required Flow Rate Estimation" This LER was reviewed in NRC Rpt 50-424/87-44 and require:
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verification of the corrective actions.
The inspector reviewed procedure 34226-C and the training attendance sheets.
The following item is identified:
50-424/ LIV 87-60-02
" Failure. to Perform required TS Surveillance to Verify compliance with TS 3.3.3.10 LER87-01" (b) *50-424/87-02, Rev 0 " Potential Failure of MSIV's to Close Following Small Steam Line Break" This LER was reviewed in NRC Rpt 50-424/87-44 and required verification of the corrective actions.
The inspection reviewed the vendor qualification report dated 3-20-87. This report documents that the main steam isolation valves which were supplied can remain in the open position for approximately 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> while exposed to a 320 degree F environment and retain the capability of closing and stopping steam flow in the system. The inspector noted that the test configuration included the relief valve (4100 psi) and that the hydraulic pressure reached only 3950 psi during the test. DCR 87 VIE 0030 was also reviewed. No further corrective actions are required as a result of the test report.
(c) *50-424/87-03, Rev 0 " Restriction of Pipe Movement with Incorrect Penetration Sealant Material" This LER was reviewed in NRC Rpt 50-424/87-44 and corrective action was verified during-the course of the event. The inspector nas no further questions.
(d) *50-424/87-04, Rev 0 " Containment Isolation Actuations Caused by Faulty Circuit Board" This LER was reviewed in NRC Rpt ' 50-424/87-44.
Corrective action was verified regarding the repair of the faulty circuit during the course of the event.
The inspector verified that a new annunciator has been added and 17006-1 response procedure changed.
In addition the inspector noted that the radiation monitors have been removed as an input to containment isolation.
(e) *50-424/87-06, Rev 0 "ESF Actuation of Auxiliary Feedwater Due to Inadvertent Trip of the Main Feedwater Pumps" This LER was reviewed in NRC Rpt 50-424/87-44 and corrective action was verified during the course of the event.
The inspector notes'that a further corrective action has oeen the practice of removing the control fuses to the actuation circuit for AFW.
This practice has resulted in LER 87-36
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when the wrong fuses were pulled.
( f) *50-424/87-07, Rev
- 0. "ESF Actuation Caused by Steam Generator Water Level"; *50-424/87-09, Rev 0 "ESF Actuation Caused by Adjustments to Steam Generator Level Control
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L 10 Systems"; *50-424/87-10, Rev 0 "RPS Actuation Caused by Adjustments to Steam Generator Level Control Systems";
"50-424/87-12, Rev 0 " Reactor Trip Due to Feedwater Control Problems Following Generator / Turbine Trip": *50-424/87-14, Rev 0 " Steam Generator High Level Results In Reactor Trip";
- *50-424/87-18, Rev 0 " Reactor Trip Caused by Faulty Bistable Circuit Board"; "50-424/87-24, Rev 0 " Procedure Inadecuacy Causes Auxiliary Feedwater Actuation";
- 50-424/87-25, Rev 0 " Reactor Trip Due to Startup Test Procedure Inadequacy"; *50-424/87-27, Rev 0 " Reactor Trip Caused by Inadvertent Closure of MSIV During Maintenance";
- 50-424/87-30, Rev 0 " Lightning Causes Reactor Trip Due to Incorrectly Grounded Current. Transformer"; *50-424/87-31, Rev 0 " Auxiliary Feedwater System Actuation During Startup
. Test Due to Procedure Inadequacy"; *50-424/87-34, Rev 0
" Reactor Trip Due to Failure - of Main Feedwater - Pump Discharge Check Valve"; *$0-424/87-35, Rev 0 " Faulty Main
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Feedwater Pump Turbine Hydraulic Tubing Connection Leads to Reactor Trip"; *50-424/87-36, Rev 0 " Auxiliary Feedwater Actuation Circuitry Inoperable Due to Personnel Error";
- 50-424/87-39, Rev 0 " Pressure Transmitter Failure Causes ESF Actuation on Steam Ge'nerator Hi-Hi Water Level";
- 50-424/87-41 Rev 0 " Reactor Trip Due to Improperly Calibrated Field Current Transducers"; *50-424/87-50, Rev 0 "Reactur Trip Caused.by Instrument Technician's Error".
These LERs were reviewed in NRC Rpt 50-424/87-38 and NRC L
Rpt 50-424/87-44 with corrective action verified during the L
course of the events.
Additional NRC concerns were l'
addressed in several management meetings regarding the control of Steam Generator water level.
Improved system performance resulted from increased operator experience and additional system tuning.
(g) *50-424/87-11, Rev 0 " Trip due to Lo-Lo Steam Generator Level" This LER was reviewed in NRC Rpt 50-424/87-44. The inspector noted that the corrective actions included temporary markings on the site glass and an engineering evaluation to determine further correction action.
The inspector questioned the final status of these two actions and was informed that no further actions were necessary.
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(h) *50-424/87-13, Rev 0 "Feedwater System Valve Malfunctions I
Result in Reactor Trip" This LER was reviewed in NRC Rpt 50-424/87-44 and at the time of the event. MWO 1-87-4987 was reviewed to verify proper reassembly.
LER 87-34 describes a repeat failure of the same check valve and describes further corrective action.
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l 11 (1) "50-424/87-15, Rev 0 " Inadvertent Steam Dump Operation Results in ESF Actuation" This LER was reviewed in NRC Rpt i
50-424/87-44 and at the time of the event. Training was 1
verified regarding the connection of test racks.
The l
inspector noted to the licensee that the LER implies that the steam' header pressure control loop was tested after tne event to ensure its proper operation was part of the corrective action, when in fact the only testing was as part of the power ascension test phase.
The licensee has not been responsive in revising the LER.
(j) *50-424/87-19, Rev 0 " Control Room Isolation Due to Signal From Toxic Gas Monitors"; "50-424/87-28, Rev 0 " Control Room Isolations Caused by Spuricus Signals From Toxic Gas Monitor" procedure 24537-1 and 24538-1 were reviewed to verify that monthly calibration checks were implemented.
It was noted that the lieansee is not required t.o have operable monitors since et:arine is removed from the site.
i The licensee is pursuing a TS change to raise the setpoint j
from 2 to 5 ppm to eliminate spurious actuations and then i
return chlorine onsite i
(k) 50-424/87-21, Rev 0 " Control Room Isolation Initiated b*y Radiation Monitor Loss of Power" The final corrective actions for this problem will be discussed along with the l
resolution of LER87-05.
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(1) *50-424/87-23, Rev 0 "RHR System Minimum Flow Requirement Potentially Not Met Due to Partially Closed Valves" Thfs LER was raviewed in NRC Rpt 50-424/87-31 and resulted in L
the identification of a Severity Level III Violation j
50-424/87-31-02.
Procedure 14460-1 was verified to have the changes and the preventive maintenance sheets indicate the calibration frequency to be every six months.
The
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corrective MW0s were also reviewed l
(m) *50-424/87-32, Rev 0 " Operator E*rc. Leads to a Reactor Trip on Source Range High Flux" Procedure 12003-1 was reviewed to verify the requirement for a ICRR plot and a rer-tor i
engineer.
Procedure 14940-1 was reviewed for to m ify incorporation of correct boron' worth and that the procedure will be performed by a reactor engineer. The training plan j
and simulator changes were reviewed.
(n) *50-424/87-33 Rev 0 " Reactor Trip on Steam Generator Lo-Lo Level While Transferring Feedwater Flow" Procedure 12004-1 was reviewed to verify that the correct power levels were indicated for transferring from the Bypass Feedwater regulating valve to the Main Feedwater regulating valve.
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(o) *50-424/87-37, Rev 0
" Failure to Meet Technical Specification Action Statement Due to Procedural Inadequacy" Procedure- 00150-C was reviewed to verify the additional guidance was incorporated.
The inspector interviewed the NSSS engineering supervisor to determi'ne the results of the LLRT performed during the outage. The results indicated that while degradation was noted the valve was within.the acceptance criteria.
The inspector determined that no actual TS violation had occurred since the valve was inoperable due to the potential that the leakage was high.
This event served in identifying a procedural system weakness.
(p) "50-424/87-38, Rev 0 " Manual Nactor Trips Due To Overly Conservative Annunciator Res anse Procedure" Procedure 17010-1 was reviewed to verify that the response procedure has been revised to place DRPI in the Data A or Data B to regain rod position indication prior to a manual trip.
(q) "50-424/87-42, Rev 0 " Boron Concentration Exceeds Tech.
Spec. Limiting Condition of Operation Time Limit" The tickler sheet was reviewed to show the correct TS limits.
The maerandum regarding surveillances was also.eviewed.
This item is identified as follows:
50-424/ LIV 87-60-03 " Failure to Adequately Perform required TS Surveillance to Verify compliance with TS 3.1.2.6.b -
4 LER87-42" (r) "50-424/87-43, Rev 0 " Improper Performance of Containment Pressure Surveillance Due to Personnel Error" Procedure 14000-1 was reviewed to verify that the computer point was included in the procedure.
This item is identified as l
follows:
j 50-424/ LIV 87-60-04 " Failure to Adequately Perform required TS Surveillance to Verify sempliance with TS 3.6.1.4 LER87-43" (s) 50-424/87-46, Rev 0 " Waste Gas Decay Tank Not Sampled Within Technical Specifications lime Limit" The memorandum regarding surveillances was reviewed.
Corrective actions include the establishment of fixed time.
This item was identified in NRC report 50-424/87-49 as an LIV.
l (t) "50-424/87-57, Rev 1 " Procedure Deficiency Results in Failure to Trip Overtemperature Delta T Reactor Trip Bistable". This LER describes an event which occurred on August 8,1987 when the shif t f ailed to place one of four
13 required bistables in trip.
The error was identified on August 9,1987 during a control panel walkdown. The root cause was a procedural deficiency in specifying the correct bistables to trip..The inspector noted that the failure-mode consisted of the pressure instrument drifting high about 40 psi and not a total failure high.
At the inspectors request engineering performed a calculation to show the effect that this pressure drift would have on the setpoint.
This calculation showed that even with this error the setpoint was within the 6.6% total allowance.
The procedure was reviewed and the corrective actions nave i
been completed. The inspector also noted that the LER was submitted late due to an improper review of the deficiency card.
Both items above represent violations of NRC requirements where the licensee has met the criteria for no citation.
To track these items the following are identified:
50-424/ LIV 87-60-05 " Failure to Place the OTDT Trip Bistables in the Trip Condition per TS 3.3.1 Item 7 - LER 87-57" and 43-424/ LIV 87-60-06 " Failure to Submit an LER Within 30 C After The Discovery of the Event per 10 CFR 50.73(a)(1) - LER 87-57" 5.
Management Meetings (303028)
On October 21, 1987, an enforcement conference was held to discuss the results of NRC report 50-424/87-56.
On November 9,1987, a site tour was given to the Director, Office of Nuclear Re&ctor Regulation (NRR), Thomas Murley and the Associate Director for Inspection & Technical Assessment, Richard Starostocki by the resident inspectors.
Following the tour, two meetings were conducted with the licensee. The first meeting was held with the Unit 1 operations personnel and the second meeting was held with the Unit 2 construction personnel.
On November 10, 1987, the fourth onsite meeting with the licensee was held i
regarding the performance of the unit.
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