ML20154L643
| ML20154L643 | |
| Person / Time | |
|---|---|
| Site: | Sequoyah  |
| Issue date: | 09/12/1988 |
| From: | Branch M, Jenison K NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20154L630 | List: |
| References | |
| 50-327-88-35, 50-328-88-35, TAC-R00398, TAC-R00399, TAC-R398, TAC-R399, NUDOCS 8809260237 | |
| Download: ML20154L643 (18) | |
See also: IR 05000327/1988035
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UNITED STATES
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NUCLEAR REGULATORY COMMISSION
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e
REGION 11
101 MARIETTA ST., N.W.
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ATLANTA. GEORGIA 30323
Reports Nos.:
50-327/88-35 and 50-328/88-35
Licensee:
Tennessee Valley Authority
6N 38A Lookout Place
1101 Market Street
Chattanooga, TN 37402-2801
Docket Nos.:
50-327 and 50-328
License Nos.:
OPR-77 and DPR-79
Facility Name:
Sequoyah Units 1 and 2
Inspection Conducted:
July 11-15 and August 22-23, 1988
Inspectors: N
b
M iL l9er
M. W. Branch, Team Leader
Date 51gned'
Team Members:
P. T. Burnett
E. F. Goodwin
Approved by: [ w 2 7 W I _ E
Se odl2.1979
K/ M. Jenisorf,Kcting Chief
Date Signdd
Pro 1ectsSectIon1
DivisionofTVAProjects
SUMMARY
Scope:
This special, identified Sequoyah Unit 2 excessive post trip cooldown
announced inspection reviewed conditions surrounding
the recently
condition and its affect on end of life shutdown margin.
The
inspection was conducted on site and at the corporate office to
independently assess the event and to evaluate the the licensee near
term and long term corrective actions.
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Results:
The inspection determined that the licensee had failed on numerous
occasions to take effective corrective action necessary to maintain
the plant as described in the FSAR, which is the reference document
for many activities, including reload core design.
Specifically
the excessive cooldown following a reactor trip had
beenidentifledin1982aspartofthestartuptestprogram.However,
) roper and adequate corrective action was not taken at that time.
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iad the licensee taken and documented effective corrective action, as
required by 10 CFR Part 50 Appendix B, Criterion ' Vi, the current
A
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problem of reduced shutdown margin would not have occurred.
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8809260237 880912
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Subsequent to the initial failure to take adequate corrective action,
the problem was not identified and pursued as a discrepancy during
any of the numerous post-trip reviews or in the reload licensing
interface process. The post-trip reviews did not adequately compare
the actual plant performance with the bases in the FSAR.
Through the
reload licensing checklist, the licensee was specifically requested
to reconfirm for each reload cycle that the FSAR RCS temperature
value was still valid.
Additionally, the standard Westinghouse Generic Emergency Response
Guideline procedure ES-0.1 was modified during procedure development
by TVA without recognizing that the intent of the procedure was to
quickly establish stable RCS temperatures at or above the no-load
value to preserve shutdown margin.
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The following Violations were identified:
Failure to take adequate corrective action when the excessive
cooldown discrepancy was first identified (Paragraph 4);
followed by subsequent failures to identify or take adequate
corrective actions during the post trip review process
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(Paragraph 7.a), as well as the 10CFR 50.59 core reload analysis
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(Paragraph 7.b), and the emergency procedure implementation
process (Paragraph 7.c).
The following Unresnived Item was identified:
Applicability of 10 CFR Part 21 requirements (Paragraph 8.b)
Completion of corrective actions to assure resolution of the
excessive cooldown encroachment on shutdown margin at end of life for
Unit 1 is a restart item.
Note:
A list of the acronyms and abbreviations used in this report
is found in Paragraph 10.
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REPORT DETAILS
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1.
Persons Contacted
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Licensee Employees
- J. Bynum, Vice President, Nuclear Power Production
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- G. Gault, Reactor Engineerin
- J. Lemons, Nuclear Engineer,g Supervisor
Nuclear fuels
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- J. Robertson, Manager, Nuclear fuels
- B. Schofield, Licensing Engineer
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- S. Smith, Plant Manager
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Westinghouse Employees
Nancy Campbell WestinghousefuelsProjectMana
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Noel Pogorzelski, Commercial Nuclear Fuel Divis$eron Core Engineer
Other licensee employees contacted included engineers, operators, and
office personnel,
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NRC Personnel
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0 Fieno, Section Leader, Reactor Systems Branch NRR
H.Richings, Engineer,ReactorSystemsBranch,NkR
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- Attended exit interview
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2.
InspectionObjectives
The Unit 2, Cycle 3 core is the first at Sequoyah to use low-leakage
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design, which has significant safety and economic benefits.
However, one
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consequence of that design was that much of the excess shutdown margin
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present in the earlier cores was lost.
Near the end of Cycle 3, the
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design shutdown margin values approached the minimum value required by TS.
Ouring the series of reactor trips which followed the restart of Sequoyah
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Unit 2, it was identified that the post-trip RCS cooldown exceeded the
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design average coolant temperatures presented in the FSAR and used in the
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accident analyses.
The affect of these excessive cooldowns was to
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decrease the available reactor shutdown margin from the values assumed in
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the cycle design and safety analysir,.
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The scope of this inspection was to evaluate information surrounding the
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discovery and proposed near and long term corrective actions associated
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with the adequacy of shutdown margin for Units 1 and 2, subsequent to
Thefollowingobjectiveswereestablished:
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To summarize the chronology of the events leading up to the
identification of the problem, and the near term actions taken to
address the problem (Paragraph 3)
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To compare the plant response in the area of RCS cooldown following
reactor trips to the expected response documented in the FSAR and the
PLS documents (Paragraph 4)
To independently verify adequate shutdown margin for several of the
p(Paragraph 5)revious reactor trips associated with the Unit 2 Cycle 3 core
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To review with Westinghouse personnel the assumptions and bases of
the core reload analyses relative to RCS temperature, and to ensure
that the margins used in the main steam line break analysis are
preserved for continued plant operation (Paragraph 6)
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To evaluate the adequacy of the post-trip review process for
identifying post-trip plant performance which deviates from the FSAR
(Paragraph 7.a)
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To evaluate the TVA design controls and vendor interface associated
with the core reload licensing (Paragraph 7.b)
To evaluate the ade.quacy of the 10 CFR, Part 50.59 safety evaluations
for core reloads with respect to this problem (Paragraph 7.b)
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To determine the reportability of the discovery of the shutdown
margin problem under 10 CFR Parts 50.72 and 50.73 (Paragraph 8.a)
To review the fuel service and design analysis contract to ensure
that 10 CFR, Part 21 requirements are implemented (Paragraph 8.b)
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To review the proposed near term corrective actions (compensatory
measures), which involve modification to the standard Westinghouse
emergency procedure to instruct the operator to emergency borate if
RCS temperature fails below prescribed values and to review the
Westinghouse proposed modifications to the shutdown margin procedure
to ensure that they )are properly reflected in the TVA procedure
(Paragraphs 6 and 9.a
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To discuss with TVA long term corrective action to address plant
cooldown subsequent to reactor trips (Paragraph 9.b)
3.
Chronology of Events
The following is a brief chronological summary of the events and
discussions which led to the discovery and the near term corrective
actions for the shutdown margin problem.
The chronology is based on
information prepared by the licensee.
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5/19/88 - Unit 2 tripped from 72 percent power and cooled to a T
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of approximately 516?F.
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5/20/88 - The NRC resident ins 3ector questioned the excessive
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cooldown associated wi;h the 5/19/88 reactor trip.
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5/23/88 - Unit 2 tri ped from 70 percent power and cooled to a T
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approxfmately512?F.
ave
of
The SI-38 Shutdown Margin
calculated following this trip indS
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that adequate
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shutdown margin was maintained dun
RCS cooldown,
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with an excess margin of 10 ppi
)ove the
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requirement.
6/6/88 - Unit 2 tripped from 98 percut .'t
, and cooled to
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T
of approximately
- ?F.
FAC resident
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iM$ectorquestionedthe
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6/13/88 - The licensee initially id-
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,ue of shutdown
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margin reduction for post-tt,p
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6/14/88 - The ifcensee initiated CAQR %,. ..ch75 to document the
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shutdown margin probleu, and discussed the core design with
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6/17/88 - Westinghouse confirmed by analysis that SOM was not
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violated in the 6/6/88 reactor trip.
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Westinghouse transmitted the minimum allowable RCS
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temperature (519?F) for maintaining the required SOM
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following a reactor trip from 70 percent reactor power
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(Reference 88TV*-G-0049).
The specified temperature was
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based on a reevaluation of the conservatisms used in the
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original calculations.
The 70 percent
aower level was
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based on plans to extend Cycle 3 operat:on into January
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1989,
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The Reactor Fuels and Analysis Branch issued boration
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volumes required for the maintenance of SOM following
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reactor trips with cooling below 520 degrees F and an
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initialpowerlevelof70 percent (L32880617901).
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6/18/88 - ES-0.1, Emergency Procedure for Reactor Trip Response, ions.
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was
revised to ensure com)11ance with technical specificat
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Revision 3 incorpora';ed a recommendstion that auxiliary
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feedwater flow be limited to maintain RCS temperature above
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520?F or a manual boration of the RCS be initiated.
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6/19/88 - NRC discovers and informs plant management that PLS has
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language to the effect that plant cooldowns should be
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restricted to preserve shutdown margin.
TVA then notifies
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NRC residents that ES-01 had been changed to compensate for
cooldown problems.
6/21/88 - P0RS initiated PRO 2-88-178 for possible reporting of the
shutdown margin problem,
6/27/88 - Westinghouse transmitted allowable cooldown temperatures
following reactor trips from 80, 90, and 100 percent
reactorpower(88TV*-G-5556).
6/29/88 - Westinghouse RF&A transmitted boration volumes required for
90, and 100 percent (L32 880629 904) power levels of 70, 80,
the maintenance of SOM from pretrip
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At the time of the inspection, Revision 4 to ES-0,1, providing guidance
for full power operation through end of cycle based on Westinghouse data
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(88TV*-G-0057), was in draft form.
Emphasis was being placed on
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maintenance program activities to repair steam leaks and help mitigate the
excessive cooldown.
Longer term options for mitigating post trip cooldown
and addressing SOM requirements were being pursued.
4.
Comparison of Actual Plant Cooldown Response to Oesign Response
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The inspectors evaluated the cooldown response of Unit 2 to several
reactor trips, and compared these responses to expected system design
responses contained in the FSAR.
The post trip cooldown problem applied
to both units.
However, for purposes of this inspection, the inspectors
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concentrated on specific data associated with Unit 2.
The Sequoyah F / states that accident analyses of the plant are based on
a no-load averar, RCS temperature of 547'F following a reactor trip.
Section 7.7, ent'tled Control Systems, and Section 15.1, entitled Accident
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Analysis, Normal Operations and Operational Transients, both indicate that
the control system are designed and groomed to maintain a post trip
no-lead 7
of 547 F.
Specifically, Section 7.7.1 states, "The steam
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dumpfeedNfercontrolsystemsaredesignedtopreventtheaveragecoolant
temperatore frem failing below the programmed nc-load temperature
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following the trip to ensure adequate reactivity shutdown margin".
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The magnitude of th, excessive RCS cooldowns for Unit 2 Cycle 3 are
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demonstrated by the data below, which was provided by the licensee:
Trip #
Date
Burnup
Power
M
Post Trip T
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48
12/29/84
45
15%
Equilibrium
510.0'F
49
1/12/85
371
99.4%
513.8'F
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50
1/14/85
385
32.4%
530.0'F
51
2/15/85
1575
99.4%
Equilibrium
506.3'F
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52
2/17/85
1582
30.2%
534.5'F
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5/3/85
4401
99.5%
Equilibrium
525.5'F
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Trip #
Date
Burnup
Power
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Post Trip T
(cont'd)
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5/22/85
4706
99.5%
Equilibrium
518.8?F
55,56 and 57 Non-Power Trips
58
5/19/88
8196
72%
516.3?F
59
5/23/88
8227
70%
512.4?F
60
6/6/88
8669
97.8%
Equilibrium
527.1?F
61
6/8/88
8677
12.4%
522.3'F
62
6/9/88
8677
19.7%
511.9?F
As can be seen from the above data the alant typically cools to an average
temperature of approximately 520?F fol'owing a reactor trip. The standard
Westinghouse design for control systems indicate that the control systems
should be able to naintain RCS temperature at or near the no-load value of
547?F.
The inspector reviewed the Sequoyah design specifics to determine if the
control systems had a different control band from the standard
Westinghouse design. The PLS document provided to TVA by Westinghouse
during construction indicated that the T
control system could control
temperature within a 4?F band. The PLS alW contained Precaution #7 which
indicated that RCS temperature must be monitored and if T
was not being
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properly controlled following a reactor trip the operatoY'should reduce
feed water in order to preserve shutdown margin.
The licensee has maintained that the plant has always experienced the
magnitude of cooldowns typical of those discussed above. To establish a
starting point for when these excessive cooldowns began the inspector
requested and was orovided a copy % reactor power (SU-9.4A).of the initial st
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response following a trip from 100
This test
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was performed in May 1982.
The inspector's review of this test fridicated
that a test deficiency (2-9.4A-1) was written against step 6.8.3 which
required that T
steady oct at or above no load T
without manual
intervention on Y$edwater flow. The test deficiency Wdicated that the
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control system could control no better than 12?F below the no-load value.
The test procedure was accepted by the PORC and annotated to the effect
that the deficiency was acceptable since there was no mandatory acceptance
criterion which it failed to meet. Additionally, the deficiency was later
reevaluated as still being acceptable based on the fact that modifications
to the main feed system to require a feed pump trip whenever a feed
isolation occurred were complete. However, no retests were performed.
The
fact that the excessive cooldowns continued af te? the modifications were
complete, and that no actions were taken to address this discrepancy
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between actual plant response and the values in the FSAR, indicate that
the corrective action for the problem, 4hich was identified through
testing, was ineffective.
Although 1'
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identified that plant
operations were not as described in ths
no written safety evaluation
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was performed as required by 10 CFR Part
.69 (See paragraph 7.b).
No
actions were initiated to update the FSAR as required by 10 CFR Part 50.71
(e).
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10 CFR, Part 50 Appendix "B" Criterion XVI, Corrective Action, requires
that measures be established to assure that conditions adverse to quality,
such as failures, malfunctions, deficiencies, deviations, defective
material and equipment, and nonconformances are promptly identified and
corrected.
Contrary to this requirement, the evaluation and correction of
the test deficiency associated with the ability to control RCS temperature
following a reactor trip was improper and ineffective, in that, system
performance stated in the FSAR was not obtained resulting in the reduction
of safety margin associated with reactor shutdown margin.
This is
identified as the first example of Violation 327,328/88-35-01.
5.
Independent Calculation of Shutdown Margin
NRC inspectors independently calculated shutdown margin for several of the
previous reactor trips on Unit 2 Cycle 3 to verify that adeouate margin
had been maintained.
The NRC requested the shutdown margin data for the five trips that
occurred since initial Unit 2 restart.
These trips were identified by the
licensee as trip numbers 58 thru 62.
The licensee was only able to
provided data for trips 58, 59 and 60 prior to the completion of the
inspection, as they indicated data for the other trips were in the
duplication process.
Using the licensee's procedure (SI-38},licensee. pectors verified the
the ins
shutdown margin calculations of the
The inspectors'
calculations were in general agreement with the licensee results.
The
inspectors then calculated the shutdown margin for trip number 60, which
occurred on June 6,1988 from 98 percent power, using the worst case RCS
temperature (the lowest RCS temperature reached during the transient). The
results of this calculation indicated that, by using the worst case
temperature and the revision of !I-38 in effect at the time of the trfp,
the TS required shutdown margin of 1.6 % delta k/k was not maintained.
This calculation indicated that the boron concentration required to
3 reserve the TS SOM requirement would be 497.7 ppm, whereas the actual
)oron concentration was 442 ppm.
However, as stated in the chronology
listed earlier, the licensee had Westinghouse perform a more refined and
precise calculation of the actual shutdown margin at the lowest RCS
temperature.
The results of this calculation, transmitted 6/17/88
(88*-G-0049), indicated that the actual shutdown margin was greater than
that requirsd by TS.
The licensee was requested to calculate, using the lowest RCS temperatures
reached, the shutdown margins for trips 61 and 62.
Both of these
calculations indicated that the TS shutdown margin values were met.
Thus, the inspectors confirmed that adequate shutdown margin had been
maintained throughout Unit 2 cycle 3 operation.
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It should be noted that the surveillance requirements for TS 3.1.1.2 do
not require an immediate determination of shutdown margin following a
reactor tri) and in no case does the TS require the lowest RCS temt'rature
be used.
Fowever, the TS requires that the shutdown margin be pr
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at all times.
Additionally, the Westinghouse developed emerge,s
procedure for actions required af ter a reactor trip recuires that a
shutdown margin calculation be performed as supplementa'
action. The
licensee's position is that they perform the shutdown margin calculation
for the actual conditions that exist at the time of performance and that
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this method is consistent with other utilities that were polled.
6.
Discussions of Shutdown Margin Design Assumptions with Westinghouse
The inspectors reviewed with Westinghouse personnel the assumptions and
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bases for the Sequoyah reload analyses ta determine whether adequate
shutdown marg'n exists on Unit 2 through the remainder of the current
cycle.
The Westinghouse design and analysis provided for the 1.6% delta k/k
shutdown margin required by TS to be preserved at end ,f life, when the
core was most vulnerable to the steam-line-break accident.
The FOR for
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Sequoyah described plant behavior post-trip as a cooldown to the r.., load
average coolant temperature of 547* F.
The Westinghouse analysis
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accounted for instrument errors by assuming the plant was operating 2
degrees F above full-load average coolant temperature immediately prior to
the trip and cooled to 2 degrees below no-load average temperature
following the trip, but the analysis did not provide any margin for any
actual cooldown below 547*F.
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Once Westinghouse was informed that the actual post-trip cooldowns were
excessive, to as low as 510* F in one loop, they perfermed a bounding
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calculation.
That calculation confirmed that none of '.he cycle 3 trips
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cnd cooldowns through June 6,1988 had reduced shutdowr margin below the
limit.
However, calculations for E0C conditions, whe i the moderator
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temperature coefficient is most negative, showed thtt under some
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conditions trips from 100% RTP would lead tv insufficient shutdown margin
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if the cooldown was to 544 degrees F or less.
Even trin from 100% RTP at
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nominal conditions, equilibeium xenon and 0 bank above 200 step withdrawn,
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would lead to reduction in shutdown margin if the RCS :ooled to less than
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532' F.
Since the cooldown for recent trips has bee; e
.<>.t 520* F, it
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was clear that additional action was necessary.
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Past practice by Westinghouse has been to reduce calce
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worth at any core condition by 10% to account for obser-
differences
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between predicted and measured control rod worths at L .
However,
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Westinghouse has justified, per approved topical report WCAP 9217, a
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reduction of the factor of conservatism to 7%.
That recalculation of
control rod worth would provide some additional shutdown margin. However,
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FSAR Table 4.3.2-3 currently shows shutdown margin to be calculated using
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10% calculated rod worth reduction.
Discussions with NRR Reactor Systems
Branch personnel revealed that they had no technical reservations about
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reducing the rod worth adjustment to 7%.
However, both they and the
inspectors objected to a further proposal by Westinghouse to reduce the
adjustment down to 5% for the current cycle of Unit 2.
The basis for that
proposal was that the 30L measurement of total control rod worth was 2%
greater than calculated.
However, that test confirms the adequacy of the
rod-worth-calculation model at BOL, it does not guarantee that the EOC
rod-worth calculation would also be an underestiniste.
At the time of the inspection, Unit 2 was being operated at 70% RTP both
to extend core life and to reduce the cooldown encroachment on shutdown
margin.
Worst case, rods at the insertion limit and transient xenon
conditions, calculations by Westinghouse showed that a post-trip
cooldown from 70% RTP to 519' F is acceptable.
addressed the possible adverse effects of "LONG TERM LOW POWER OPERATION
OF PRESSURIZED WATER REACTORS."
Westinghouse personnel stated that
specific guidance on that concern had been provided to TVA and that the
plant was currently operatinc
in conformance with the guidance.
The
essential element of the gu< dance is to naintain the axial power
distribution at reduced power consistent with that expected at full power.
That is accomplished by operating with control rods partially inserted at
reduced power.
7.
Evaluation of Design Controls and Vendor Interface
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Af ter the excessive cooldown was identified as a test deficiency in 1982,
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but was not adequately corrected, there were a number of additional
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opportunities to identify the problem.
The inspectors reviewed activities
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related to design and safety review processes to detettaine why the
excessive cooldown and the resulting effect on shutdown margin were not
identified through other available mechanisms,
a.
Post Trip Review Process
The inspectors questioned why the discrepancy between the actual RCS
temperature and the FSAR values following a reactor trip was not
identified during any of the post-trip reviews which had been
performed.
The inspectors reviewed the post trlp review procedure AI-18, File 18
in order to determine if there was a requirement to evaluate core
performance (i.e. , shutdown margin) for the lowest temperature
reached during the transient. This procedure does not specifically
require that actual core or olant performance be evaluated against
specific FSAR transient analysis requirements. It has only a simple
(yes/no) statement that 611 designated parameters were within
expected limits. Had the post trip procedure been more detailed and
required the evaluation of worst case shutdown margin or compared
actual post trip parameters with FSAR values, the fact that control
systems were incapable of satisfying FSAR requirements regarding
lowest T
following a reactor trip and the calculated violation of
shutdewn drgin requirements would have been identified earlier.
The
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licensee was requested to evaluate the procedure for calculating port-trip
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shutdown margin and determine if required results can be achieved
with current procedure detail.
The surpose of the post trip review is to identify and correct
cond'tions that are not acceptable and affect plant expected response
to trip conditions. The failure of recent post trips reviews
performed subsequent to the May 19 and 23,1988 and June 6, 1988
reactor trips a well as post tri,; reviews performed prior to the
August 1985 shutdown
to identify and correct the plant cooldown
condition is identifled as the second example of violation 327,
328/88-35-01 for ineffective corrective action .
b.
Vendor Interface and 50.59 Safety Evaluation for Reload Cycles
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To determine why the cooldown discrepancy was not uncovered during
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the design review for the core reloads, the inspectors examined the
interface between the licensee and the vendor, and reviewed the
10 CFR 50.59 safety evaluation performed for Unit 2 Cycle 3 and Unit
1 Cycle 4.
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Although not totally within the scope of this inspection, the
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Westinghouse interface efforts supporting a core reload analysis were
briefly discussed with the Westinghouse Fuel Project Manager and a
Westinghouse core engineer.
The Westinghouse personnel presented the
following scope and timetable for the reload design interface
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process:
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FORMAL TVA/ WESTINGHOUSE COMMUNICATIONS
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RELOAD DESIGN PROCESS
MONTHS PRIOR TO STARTUP
A. INITIALIZATION PHASE
Reload Safety and Licensing Checklist
18
Energy Requirements
Preliminary Loading Pattern
16
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Design Initialization Meeting
14
Design Scheidule
13
8. CORE MANAGEMENT PHASE
Loading Pattern Established
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HONTHS PRIOR TO STARTUP
(cont'd)
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C. SAFETY ANALYSIS PHASE
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Reload Safety Analysis Checklist
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Reload Saf,ety Evaluation
4
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D. OPERATIONS INFORMATION
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Nuclear Design Report
and Operations Data
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The reload safety and licensing chacklist was described by Westinghouse as
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the vehicle where plant specific performance parameters are transmitted
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between Westinghouse and TVA for the purpose of validation of data.
The
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following is an excerpt from the Sequoyah Unit 2 Cycle 3 Westin
Reload Safety and Licensing Checklist, Revision 0, dated 7/21/83. ghouse
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The performance characteristics of plant components or safety systems
assumed in prior safety analyses are important input to the safety
e
analyses for the next cycle.
Unless otherwise advised, the
performance characteristics found in the following documents will be
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assumed for the next cycle's licensing effort.
The documents
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revisions must be consistent with the date of issuance of the
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completed checklist.
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(1) The original Plant Safety Analysis Report
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(2) Loss of Coolant Accident Submittals
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(3) Fuel densification submittalc (if any)
(4) Reload Safety Evaluation Reports
(5) Technical Specifications (approved or submitted)
(6) Any other special analyses such as Anticipated Transient Without
Trip (ATWT) analyses, analyses of the effect of a modified
system, etc. unless addressed in your response. Westinghouse
must assume that the only changes in core characteristics for the
reload are those found in the design of the reload core.
Section II-C
Thermal Hydraulics
(a) Change in operating pressare - none
(b) Change in operation temperature - none
T in
--- 547 degrees F
T ave
--- 547 degrees F to 578 degrees F
Delta T --- 60.3 degrees F
.
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Thus, the Westinghouse reload licensing checklist specifically states that
i
the FSAR values, including no-load T
will be assumed in the reload
design analysis unless Westinghouse $, notified otherwise by TVA.
A
Therefore, the licensee had been formally made aware of the continuing use
of the originally assumed no-load T
design value for the reload
calculations.
AVE
The inspector evaluated the adequacy of the TVA 10 CFR Part 50.59
,
eva'uations for the Unit 2 Cycle 3 and Unit 1 Cycle 4 reloads to attempt
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to determine why the expected effects of excessive plant cooldowns were
not addressed in these safety evaluations.
USQO 84-34 for Unit 2, dated
9/7/d4, and USQD 85-20 for Unit 1, d&ted 11/1/85, contained only 2 pages
a
.
and consisted of nothing more than cover sheets with signature blocks for
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the Westinghouse Reload Safety Evaluation.
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10 CFR 50.59 allows the holder of a license to make changes in the
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facility as described in the safety analysis report without prior
commission approval unless the proposed change involves a change in the
technical specifications incorporated in the licensee or an unreviewed
,
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safety question.
A proposed change shall be tieemed to invo' 'e an
unreviewed safety question; 1) if the probability of occurrence or the
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consequences of an accident or malfunction of equipment importaat to
safety previously evaluated in the SAR be increased; or 2) if the
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possibility for an accident or malfunction of a different type than any
evaluated previoudy in the SAR may be created; or 3) if the margin of
,
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safety as defined in the basis for any technical specification is reduced,
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In performing the 10 CFR 50.59 safety evaluations for the reload cores,
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the accuracy of the post-trip cooldown values presented in the FSAR were
assumed to be correct and were not questioned, based on the assumption
that tite FSAR had been kept up to date.
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The initial failure to take adequate corrective action when the excessive
!
cooldown was initially identified including the failure to comply with
10 CFR 50.59 at that time was addressed in paragraph 4 as violation
327,328/88-35-01.
As a result of the initial failure, both Units 1 and 2
!
had been operated since licensing outside the system design described in
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the FSAR. Specifically, FSAR Section 7.7.1 required that the steam dump
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and feed water control system be designed to prevent the average coolant
temperature from falling below the program no-load temperature following a
2
reactor trip to ensure adequate reactivity shutdown margin is preserved.
The excessive cooldown constituted a change to the operation of the
facility as described in the FSAR and should have been supported by a
written safety evaluation.
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The Westinghouse analysis, based on the criteria citablished by the Reload
Safety and Licensing checklist discussed above, used as a basis a post
trigT
value of two degrees less than the no-load value of 547 F (i.c.
545 F). ave Using this post trip temperature of 545'F resulted in a
calculated EOC shutdown margin of 1.61 % delta K/k compared to the TS
required value of 1.6% delta k/k for Unit 2 and 1.64 % delta k/k compared
_ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
. _ _ _ _
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12
to the TS value of 1.6 % delta k/k for Unit 1.
Although the amount of
excess shutdown margin available at E00 tended to be lessened by the low
leakage design, this reduction in itself did not constitute an unreviewed
safety question per 10 CFR 50.59 as long as the required 1.6 % delta k/k
was maintained,
Review of corresponrience from Westinghouse to TVA reaarding this issue
i
included a June 27, 1988 letter (88TV*-G-0056) in which Westinghouse
stated that cooldown temperatures as low as 520'F might result in future
loading pattern restrictions, which would reduce the low-leakage
,
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capability, with loss of its attendant advantages.
The amount of shutdown
margin reduction associated with a post-trip temperature change from 545'F
,
!
to 520 F would reduce the EOC shutdown margin value by approximately 0.7%
delta k/k down to approximately/k.0.9 % delta k/k, which is below the
allowed TS value of 1.6% delta k
Westinghouse recommended limiting the
i
cooldown as discussed in paragraph 7c.
The adequacy of the reload cycle 10 CFR 50.59 evaluations, which did not
2
1
identify as an unreviewed safet
ouestion the decrease in E0C shutdown
margin below the value allowed
TS is identified as the third example of
violation 327,328/88-35-01 for i effective corrective action,
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The inspector reviewed the following Nuclear Fuels Procedures to determine
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if, in general, adequate design and interface controls exist:
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NFP 7.0, Control of Reload Core Design and Analysis
NFP 7.1, Organization and Interface for Reload Design and Analysis
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NFP 7.2, Reload Design Document Control
The inspector concluded, based on the procedures reviewed, that in
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general, acequate controls did exist to obtain a proper core reload
analysis.
However, the procedures reviewed were issued in 1987 rather
than 1983 when
the analysis was performed.
The manager of Nuclear
Fuels indicated that similar procedures did exist at the time the core
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reload analysis was performed.
Additionally, it should be noted that even
the current procedures will be subsequently modified to reflect the
February 1988, organizational change that made the Nuclear Fuels Division
-
a part of DNE.
c.
Emergency Procedure Review
Since part of the TVA corrective action was to modify the standard
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Westhghouse owner's group emergency operation procedure to
compensate for the excessive cooldowns, the inspector conducted a
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review of the procedure and the Westinghouse guidance. Specifically,
!
procedure ES-0.1 Recctor Trip Response was reviewed. The Westinghouse
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guideline contained language to the effect that RCS and secondary
1
plant stabilization at no-load conditions was part of the procedures
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major action goals. In fact, the logic tree for step 1 of the
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procedure shows actions required for temperature decreasing below the
no-load values as stop dumping steam followed by controlling AFW flow
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13
to maintain SG 1evel at the bottom of the level band and to isolate
the main steam line if necessary.
The TVA implementing procedure issued October 4,1984, did not
specify the Westinghouse course of action to preserve temperature at
or above the no-load value.
The TVA procedure stated that if Tave is
decreasing in an uncontrolled manner, then verify steam dumps and
valves. y PORV closed followed by closing the MSIVs and their bypass
secondar
This method does not appear to preserve the no-load
temperature and consaquantly the reactor shutdown margin.
The
inspector reviewed the TVA stap deviation for this procedure issued
subsequent to procedure implementation and determined that the TVA
basis for this deviation was that the AFW system design includes
automatic level control valves and therefor manual control of AFW is
not necessary.
This deviation does not appear to address the issue at
hand to preserve shutdown margin possibly at the expense of reducing
If, at the time of implementation of the generic guides in
Oc' Wer 1984, TVA had questioned the pui aose of the steps in the
W . inghouse procedure the excessive coulcown/ shutdown margin problem
1
may have been properly resolved at that time.
This failure to
identify and correct a nonconforming condition is identified as the
fourth exam
tive action.ple of violation 327,328/88-35-01 for ineffective correc-
8.
Reportability
a.
10 CFR 50.72 and 50.73
As indicated above, the NRC considers that the reduction in EOC
shutdown margin associated with the excessive
slant cooldown
constituted an unreviewed safety cuestion and cou d have resulted in
the plant being in a condition tha", was outside the design basis.
In
fact, the licensee's near term corrective action was to limit reactor
power to 70 percent and to change the standard Westinghouse post-trip
emergency procedure as a compensatory measure to ensure that the
plant could be operated within the design basis.
The CAQR (SQP880375) dated 6/14/88 indicated that the discovered
condition was not reportable.
The copy of PRO (2-88-178) dated
6/21/88 provided to the inspectors did not have a reportability
determination made at the time of the inspection.
However the
licensee did provide the written report , LER 328-88-030 withinthe
required 30 day period.
b.
Due to the potu ttal generic implication of the above shutdown margin
problem, the inspector reviewed the Fuels and Analysis service
contract (68p-84-TI) between TVA and Westinghouse to determine if the
requirements of 10 CFR 21 regarding vendor re:ponsibility as to
__ _ ___- - _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _
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______ _____________
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14
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notification were applicable.
The inspector determined that the
original contract dated in 1966 was issued prior to the January 6,
1978, date specified in 10 CFR 21. However, this contract has been
amended several times since Part 21 first became applicable. None of
the contract amendments contained language that the requirements of
10 CFR Part 21 apply.
The contract did however, contain language to
s
the effect that all NRC
rules and regulations both current and
'
future apply. The inspector requested that the licensee evaluate the
current contract and determine if it should be amended to
,
i
specifically state that Part 21 applies.
This item is identified as
1
unresolved item 327,328/88-35-02 pending further NRC review with the
licensee and the NRC vendor branch.
!
]
9.
Review of Corrective Measures
)
a.
Near Term Compensatory Heasures
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]
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As previously described, the licensee's near term corrective actions
i
,
included operating at a reduced power level of 70% RTP with power
t
distribution guidance provided by Westinghouse.
!
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In addition, current and proposed plant procedures require post-trip
'
emergency boration as a compensatory action to restore shutdown
margin rapidly if the cooldern is beyond power and burnup dependent
limits.
The limits and required boration were obtained from
Westinghouse, but before they were accepted and implemented they were
,
subjected to independent review and an>>1ysis by the TVA PWR Core
Design Section of the Reactor Fuel and Analysis Department.
The
!
inspectors' review of the records confirmed that TVA used independent
core performance calculations to confirm that the vendor calculations
.
gave results equivalent or conservative with respect to theirs.
The
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TVA methods have not been described in a topical report approved by
NRR, but were deemed acceptable for quality control purposes.
Finally, the TVA calculations were reviewed by independent reviewers
and the differences from Westinghouse results rationalized by a
reviewer familiar with Westinghouse methodology.
The TVA staf f
generated curves and tables of required boration as a function of
,
burnup, power level, and cooldown using a computer program written in
house.
That program is well-documented internally, and has been
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accepted by peer review.
The inssectors concluded the TVA review of both Westinghouse and
internal
calculations was satisfactory in both conduct and
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documentation.
The inspectors did express one concern with the
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procedures that have or will result from these activities.
The
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procedures will specify the volume of 21,000 ppm boric acid to be
!
injected.
At other facilities the boric acid and primary water flow
i
integrators have not shown acceptable accuracy for this purpose.
The
!
calibration and reliability of the boric acid integrator was not
!
established during this inspection.
The inspectors expressed this
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15
.
concern to the TVA staff during the inspection and to management at
the exit interview on July 14, 1988.
The inspectors discussed the licensee's calculation activities and
proposed compensatory actions and procedures with members of the NRR
Reactor Systems Branch.
The NRR staff had no criticism of either the
calculations or compensatory action for Unit 2 as described to them
by the inspectors.
The NRR staff did express reservations about
accepting similar compensatory action for Unit 1, which is faced with
the same problem at EOC, but has yet to restart after being refueled
.
during the current outage.
That reservation was forwarded to plant
management at the exit interview.
Management stated they did not
intend to restart Unit I until the basic problen of excessive
'
cool-down to an unanalyzed temperature had been resolved.
Management
r
further stated
they intended to ecmplete their determination of the
best method to limit post-trip cooldoen within 30 days,
b.
Long Term Corrective Actions
'
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As part of this inspection the inspectors planned to discuss details
of TVA's plans to minimize RCS cooldowns following reactor trips.
The licensee indicated that they are currently invrGtigating several
methods to attempt to control cooldowns. They include a chance in
i
possible modifications to D$ control to Steam Pressure contro', and
steam dump setting from T
auto-level controls associated with the
AFW system.
Details and schedule were not discussed.
10.
List of Acronyms and Abbreviations
AI
Administrative Instruction
-
-
ATWT
Anticipated Transient without Trip
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,
CAQR
Condition Adverse to Quality Report
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,
CFR
Code of Federal Regulations
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Beginning of Cycle
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>
BOL
Beginning of Life
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Division of Nuclear Energy
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End of Cycle
-
Emergency Procedure
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Final Safety Analysis Report
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LER
Licensee E,ent Report
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MSIVs -
NFP
Nuclear Fuels Procedure
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NRC
Nuclear Regulatory Commission
-
Nuclear Reactor Regulation
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Precautions Limitations and Setpoint Document
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Plant Operations Review Committee
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P0RS
Plant Operations Review Staff
-
>
Power Operated Relief Valve
-
Parts Per Million
-
t
- _ _ _ _ _ _ . _ _ _
_
_
_ _ _ _ _
_
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_ _ _ _ _
_ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
.
.-
16
.
Pressurized Water Reactor
-
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RF&A
Reactor Fuels and Analysis
-
Rated Thermal Power
-
Safety Antlysis Report
-
-
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Surveillance Instruction
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Sequoyah
-
SU
Start Up Test Procedure
-
T
AVE
Average Reactor Coolant Temperature
-
TS
Technical Specifications
-
Tennessee Vallay Authority
-
NRC Unresolved item
-
Unreviewed Safety Question
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USQD
Unreviewed Safety Question Determination
-
XE
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11.
Exit Interview
The inspection scope and findings were summarized on July 14, 1988, and
again on August 3,1988, with those persons indicated in paragra3h 1
above. The inspectors described the areas inspected and discussec
in
detail the inspection findings listed below. During the course o' the
inspection the' inspectors were provided numerous documents which the
licensee considered as proprietary. However, no proprietary material is
contained in this report. Dissenting comments were not received from the
,
!
licensee during the July 14, 1984 exit.
However, during the August 23.
1988 reexit the licensee did comment that their position was that the
shutdown margin 3roblem was licensee identified and was not prompted oy
tne NRC questioning of the excessive cooldown discussed in this report.
Item Numbsr
Description and Reference
327,328/88-35-01
Violation:
Failure to take adequate corrective
action when the excessive cooldown discrepancy
was first identified (Paragraph 4); followed by
,
subsequent failures to identify or take adequate
corrective during the post trip review process
(Faragraph 7.a), as well as the 10 CFR 50.59 core
reload analysis (Paragraph 7.b) and the emergenc
procedure implementation process (Paragraph 7.c)y
.
327,328/88-35-02
Unresoived Item:
Determine the applicabilit
10 CFR Part 21 :equirements (Paragraph 8,b) y of
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.
. _ . _