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U.S. NUCLEAR REGULATORY COMMISSION | |||
REGION III . | |||
Reports No. 50-456/92023(DRP); 50-457/92023(DRP) | |||
Docket Nos. 50-456; 50-457 Licenses No. N?F-72; NPF-77 | |||
Licensee: Commonwealth Edison Company | |||
Opus West III | |||
1400 Opus Place | |||
Downers Grove, IL 60515 | |||
Facility Name: Braidwood Station, Units 1 and 2 | |||
Inspection At: Braidwood site, Braidwood, Illir.ais | |||
Inspection Conducted: October'13 through November 30, 1992 | |||
Inspectors: S. G. Du Pont | |||
J. R. Roton | |||
G. M. Hausman | |||
Approved By: M.Farbh,C M 7!92- | |||
Reacto/ProjectsSectionlA Date | |||
' | |||
Inspection Summary- | |||
s | |||
Inspection from October 13 throuch November 30. 1992 (Reports No. 50- | |||
456/92023(DRP): 50-457/92023(DRP)) | |||
Areas Inspected: Routine,. unannounced safety inspection by the resident and | |||
regional inspectors of licensee action on previously identified items; | |||
licensee event report review; outages; radiation protection; operational | |||
safety verification; mont!.ly surveillance observation; and report review. | |||
Results: Three violations were identif':ed in one of the six areas inspected. | |||
In the remaining areas, no violations were identified. | |||
The. following is a summary of the licensee's performance during this | |||
inspection period: | |||
I | |||
L Plant Ooerations | |||
The licensee's performance in this area for this inspection period.was1 | |||
good. Shift briefings continued to provide sufficient'information for. | |||
l planned evolutions to be performed during the shift. The. inspectors | |||
have raised several questions involving operability determinations | |||
associated with the Main Steam Line Code Safety Valves.- | |||
_ | |||
l .9212300007 921218 - ' | |||
PDR ADOCK 05000456'- | |||
G PPR | |||
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4,= | |||
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Radiolooical Controls- | |||
Three violations were issued due to the licensee's failure to adequately | |||
control the addition of SF -to the steam generators (two violations) and | |||
the failure to adhere to the posting requirements' of Radiologically | |||
Controlled' Areas'. Additionally, the report discusses activities | |||
associated with safety evaluations for the SF, and a chloride excursion. | |||
One was an example of good efforts producing a detailed evaluation and | |||
-the other was an example of a failure to perform an evaluation. | |||
Safety Assessment /0uality Verification | |||
The one LER reviewed during this inspection period appears to have | |||
appropriate corrective actions to preclude similar events. The | |||
licensee's evaluation of the Unit I chloride excursion is a good example | |||
of a detailad and comprehensive safety assessment. However, the failure | |||
to conduct a similarly comprehensive evaluation for the sulfur | |||
hexafluoride addition indicates that the sensitivity to and | |||
understanding of the need for safety assessments is not uniform | |||
throughout the licensee's organization. | |||
Enaineerina and Technical Support | |||
Due to the inspectors limited review in this area, the licensee's | |||
performance was not assessed for this inspection period. | |||
Maintenance and' Surveillance | |||
The licensee's performance in maintenance and surveillance during this | |||
inspection period was good. | |||
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DETAILS- | |||
1 -. Persons Contacted , | |||
Commonwealth Edison Comnany (CECO) | |||
*K. L. Kofton,-Station Manager , | |||
G. R. Masters, Project Manager | |||
G. E. Groth, Production Superintendent | |||
D. E. O'Brien, Technical' Superintendent | |||
D. E. Cooper, Assistant Superintendent - Operations ' | |||
R. J. Legner, Services Director | |||
A. O Antonio, Nuclear Quality Program Superintendent . | |||
*R. Byers, Assistant Superintendent Work __ Planning | |||
G. Vanderheyden,. Technical Staff Supervisor | |||
S. Roth, Security Administrator | |||
K. G. Bartes, Nuclear Safety Supervisor ' | |||
A. Haeger, Regulatory Assurance Supervisor | |||
*J. Lewand, Regulatory Assurance | |||
S. Hunsader, EQ Supervisor Design Support - Nuclear Engineering | |||
K. C. Radke, Technical Staff System Engineer | |||
* Denotes those attending the exit interview conducted on November 30, | |||
1992. | |||
The inspectors also interviewed several other licensee employees. | |||
2. Licensee Action on Previously identified Items (92701. 927021 | |||
a. Ol en item | |||
LClosed 50-456/92017-02(DRP): 50-48i7 /92017-02(DRPl: Failure to | |||
Mlow Posting Requirements of a Radiologically Controlled Area. | |||
Inspection Report 92017 details the f ailure of a Radiological | |||
4 Protection Technician (RPT) to adhere to the posting requirement | |||
to conduct 'a whole body frisk prior to exiting _ a radiologically | |||
controlled area (RCA). In their followup review, the inspectors | |||
' | |||
discovered that-two weeks prior to this incident, a RPT had failed | |||
- | |||
to verify the decontamination of the 1A letdown heat exchanger | |||
room before removing the posting. As a result, the RPT and one | |||
- Electrical Maintenance Department person were contaminated when | |||
they entered the room to replace light bulbs. At:ditionally, there - | |||
has been one other incident since the open item was identified. In | |||
this incident, _two Mechanical Maintenance Department personn_el | |||
failed to adhere to the posted requirements-for entry into a RCA | |||
and were subsequently contaminated. These failures to adhere to | |||
the posted requirements for conducting work within a RCA are | |||
, | |||
violations of Braidwood Technical Specification 6.11, " Radiation | |||
Protection Program," as detailed in Braidwood Radiation Protection | |||
Procedure 1110-3, " Radiological Postings, Labels, and Controls," | |||
(50-456/92023-01(DRP); 50-457/92023-01(DRP)). | |||
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(Closed) ODen Item (456/88010-Ol(DRS):457/880ll-01(DRS)): | |||
Adequacy of fire protection for several unprotected structural | |||
steel columns and auxiliary steel attachments. The columns were | |||
located in the fuel building between V and W at coordinates 17, | |||
18, and 19, and the attachments were in the auxiliary building on | |||
column P-21 at elevation 401'-0". The~ licensee was to provide the l | |||
methodology used for selection and_ identification of the fire 1 | |||
protected structural steel components and the technical 1 | |||
justification that column P-21 met the specified fire rating. l | |||
For the culumns, the licensee stated that because of the low fire | |||
loading and the large open volume of the area, a credible fire | |||
would pose no hazard to the structural steel columns, therefore, | |||
fire proofing of the columns was not required. Justification was | |||
provided in Sargent & Lundy Engineers letter dated May 20, 1988. | |||
The columns support the slab at elevation 451' 0", a portion of | |||
which carries a fire rating. The calculated fire loading for the | |||
area, which includes an allowance for transient combustibles, is | |||
5000 Stu/ft' (Fire Protection Report, Subsection 2.3.12.1). This | |||
equates to a fire severity of under four minutes duration (NFPA- | |||
Fire Protection Handbook, Chapter 9, Section 7). Therefore, a | |||
credible fire would pose no hazard to the structural steel | |||
columns. | |||
For the auxiliary steel attachments, the licensee stated that- the | |||
additional heat transfer into the fire protected column from the | |||
unprotected auxiliary steel attachments did not degrade the fire | |||
rating for column P-21 below the specified three hour rating. | |||
Justification was supplied in Sargent & Lundy Engineers letter | |||
dated May 20, 1988. The column was protected by a fire-proof | |||
material, Pyrocrete 102 (7/8" thick), in accordance with | |||
applicable installation drawings, which designated a three hour | |||
fire rating according to Underwriters Laboratory (UL) Detail | |||
X-719. The UL rating was based on tests conducted on a W10x49 | |||
column. P-21 was a W14x342 column, which had a cross section | |||
seven times as massive as the VL tested column. The American Iron | |||
and Steel Institute (AISI) had performed extensive research and | |||
tests on a wide range of column sizes including sections which | |||
were more massive than the UL tested W10x49 column. These tests | |||
were summarized in AISI publication " Design Fire Protection for | |||
i Steel Columns," Third Edition, March 1980, which indicated that | |||
L the effective fire rating of the W14x342 column was more than | |||
t twice that for the W10x49 column. Therefore, ample margin was | |||
L provided to compensate for the additional heat input from a | |||
i potential fire due to the unprotected auxiliary steel attachments. | |||
Based upon the above, the inspectors concluded that the | |||
methodology and technical justification provided were acceptable ' | |||
and the inspectors had no further concerns. This item is closed. | |||
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4 | |||
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2: | |||
b. Unresolved Items | |||
(Closed) Unresolved Item (456/90Q19-02(DRSl: Physical separation | |||
between fuel oil overflow, supply, and vent lines associated with | |||
the opposite train emergency diesels. The licensee conducted a | |||
detailed review'and analysis ~that determined the installed piping | |||
arrangement, although, not consistent with the configuration | |||
described in the fire hazard analysis-(FHA), posed no immediate | |||
operational concerns for the opposite train diesel.. The FHA | |||
stated that all equipment, cables, and piping in the diesel | |||
generator room, the diesel oil day tank room, and the diesel oil- | |||
_ | |||
storage tank room would be associated with only one ESF division. | |||
However, physical inspections revealed that some of the diesel oil | |||
piping on each train was routed through the three rooms of the | |||
opposite train. | |||
. | |||
A detailed review of the Byron /Braidwood Fire Protection Report | |||
(FPR) was made to identify all other inconsistences related to | |||
physical separation of safety related equipment, cables, and | |||
piping. This consisted of a zone-by-zone review of the FHA | |||
(Section 2.3 of the FPR), a review of the safe shutdown analysis | |||
(Section 2.4 of the FPR), and a review of. FPR Section 3.0, which | |||
addressed conformance to the Standard Review Plan. No other | |||
inconsistences from the FPR were identified. Minor changes were | |||
completed to address the inconsistences as described in Sargent & | |||
Lundy Engineers Letter dated November 29, 1990, and'Transmittcl | |||
DIT-88-EXT-0124, " Assessment of Diesel Oil Piping Routed in | |||
Opposite Train Diesel Generator Rooms" dated February 25, 1992. | |||
The inspectors reviewed the licensee's detailed analysis and | |||
discussed the results with NRR. The discussion concluded that the | |||
licensee's analysis was acceptable and that Appendix R concerns | |||
were adequately addressed, since offsite power would be available | |||
and the diesel oil piping would remain intact during a postulated | |||
fire. . Corrective actions taken by the licensee indicated prompt. | |||
actions were performed including an expanded scope of review and | |||
analysis which included the Byron Station, as-well-as, evaluating | |||
the probability of missile affects on the opposite train diesel | |||
piping. The inspectors noted, however, that during the | |||
performance of two minor changes numerous field problem. reports | |||
(FPRs) were generated relating to interference and clearance | |||
. problems indicating that planning was not effective. The | |||
inspectors had no further concerns and considered this item | |||
closed. | |||
c. Violation | |||
(Closed) 50-456/92017-03(DRP): 50-457/92017-03(DRP): Technical | |||
Specification 6.8.1 was violated when the licensee failed to- | |||
convert Nuclear Work Requests to Temporary Alterations in | |||
accordance with Braidwood Administrative Procedure 2321-18. 'The | |||
licensee's response to-this violation was prompt and thorough. | |||
5 | |||
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_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ | |||
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Actions taken appear adequate to preclude recurrence of this or | |||
similar events. This item is closed. | |||
One violation was identified. | |||
3. Licensee Event Report (LER) Review (92700) | |||
LER: were reviewed and closed based on the following criteria: | |||
* Reportability requirements were met. | |||
* Immediate corrective actions were accomplished. | |||
* Corrective actions to prevent recurrence has been or will be | |||
initiated per technical specifications. | |||
No violations or deviations were identified. | |||
(Closed) 457/01006: Reactor Trip Due to Valve Hispositioning. This | |||
event is discussed, in detail, in Inspection Report 50-456/92020; 50- | |||
457/92020, Paragraph 4. In addition to those corrective actions | |||
previously identified, the inspector also notes the involvement of GN | |||
Venture contractor personnel in the assessment of root cause | |||
determination and corrective action. This item is closed. | |||
4. Outanes (86700) | |||
No violations or deviations were identified. | |||
. On November 3,1992, Unit 1 Main Generator was synchronized to the grid, | |||
ending the licensee's planned 66-day refueling outage seven days ahead | |||
of schedule. In addition to completing ahead of schedule, the 185.0 Rem | |||
of exposure and 88 personnel contaminations were well under the ALARA | |||
goals of 208.5 Rem and 126 personnel contaminations established for the | |||
outage. From a budgetary standpoint, initial estimates show the outage | |||
to be $1.8 million under the approved budget. The outage was - | |||
accomplished without major incidents and difficulties. The lessons- | |||
learned must be carried forward into the refueling outage for Unit 2, | |||
which begins in March 1993. | |||
M 5. Radiation Protection (937011 | |||
Two violations were identified pertaining to radiological work | |||
practices, performing safety analysis, and preparing written procedures | |||
for chemistry related evolutions. | |||
, | |||
* | |||
Addition of sulfur hexafluoride (SF.). | |||
* Evaluation of planning and implementation (SF.). | |||
* Inspectors conclusion (SF ). | |||
* Apparent violations (SF.) . | |||
* November 6,1992, chloride excursion. | |||
* Chloride excursion safety evaluation. | |||
Addition of sulfur hexafluoride (SF.) causes unexpected variations in | |||
steam generator chemistry. On November 6, 1992 the Braidwood Chemistry | |||
Department inadvertently caused a chemical variation on the Unit 1 steam | |||
6 | |||
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generators. Technicians injected a large amount, Labout 15 standard | |||
cubic; feet, of sulfur hexafluoride (SF.) gas into the condensate system. | |||
The injection of SF, was part of troubleshooting efforts on Unit 2--steam '! | |||
generators. _ Probable leakage in the Unit' 2. steam generators resulted in .i | |||
unexpected chemistry leve'is. Since the condensate system contains | |||
several connections between the units, the licensee ;uspected that these i | |||
connections were.the leakage source. | |||
' | |||
Shortly after injecting SF., the technicians noticed unexpect' ed | |||
responses in the Unit 1 steam generators' chemistry. A chemistry sample | |||
confirmed that a large variation occurred. This required entry into the | |||
action level of the station chemistry. procedures. The sample displayed | |||
elevated cation conductivity, phosphates, fluorine, and sulfonates. -The | |||
technicians determined that the SF, gas unexpectedly broke down into | |||
sulfonates and fluorine. | |||
The inspectors evaluated the planning and implementation of _the | |||
troubleshooting efforts involving injection of SF ._ Previously, SF, | |||
injection was used to find leaks in the condenser water boxes. The | |||
success of this usage influenced the chemistry department to use SF, in | |||
the condensate. | |||
- | |||
However, they did not evaluate the possibility of SF, going in the steam | |||
generators. They also did not evaluate the effects of the -steam | |||
generator water chemistry on SF, . S F, gas is only slightly soluble in | |||
water and soluble in alkaline solutions. Since water chemistry is- | |||
alkaline, SF, broke down into fluorine and sulfonates. This condition | |||
is not desirable since fluorine is corrosive to the heat transfer | |||
surfaces and sulfonates is basic. The technicians injected SF, into | |||
the condensate system without considering th'ese effects. They also | |||
injected the SF, without a procedure. | |||
The inspectors concluded that the chemistry department did not follow | |||
several procedures and requirements before and during the injection of | |||
the SF. gas. 10 CFR 50.59, " Changes, Tests,-and Experiments," requires .' | |||
performance of an evaluation of tests or experiments not described in | |||
the safety analysis report. The regulation also requires the . evaluation- | |||
for changes in the facility. _This evaluation is to determine the | |||
possibility of an u_nreviewed safety question or a change in the | |||
1 technical specification. | |||
The Braidwood safety a'nalysis report describes the methods for | |||
naintaining water chemistry in the steam generators. The addition of | |||
ammonia in the form of ammonium hydroxide, or an equivalent amine, and | |||
hydrozine to the condensate is in Section 10.3.5.1. The addition of SF, | |||
gas to_the condensate is not in-the safety analysis report. | |||
TD regulation 10 CFR 50, Appendix B, Criterion V, " Instructions, | |||
< Procedures, and Drawings,' requires that activities affecting quality | |||
shall be done by documented procedures. These. procedures shall be of a | |||
type appropriate to the circumstances of the activity and followed. | |||
Additions *,f chemicals to the condensate affects the steam generator and | |||
can affect the quality _of the heat transfer surface. | |||
7 | |||
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The Braidwood Technical Specification, Section 6, " Administrative | |||
Controls'," requires-the_ establishing _of procedures for activities | |||
prescribed in Regulatory Guide 1.33,. Quality Assdrance Program | |||
Requirements. The regulatory _ guide requires the establishing of | |||
procedures for controlling water quality. These procedures _will contain | |||
limits for concentrations of agents that are corrosive to heat transfer | |||
surfaces. The addition of SF. into the ccndensate was performed without-- | |||
establishing a procedure. | |||
The inspectors concluded that the following activities were violations | |||
of NRC requirements. The injection of SF, without a procedure was an | |||
apparent violation of Technical Specifications (50-456/92023-02(DRP); | |||
50-457/92023-02(DRP)). The failure to perform an evaluation for | |||
unreviewed safety condition is an apparent violation of 10 CFR 50.59 | |||
(50-456/92023-03(DRP); 50-457/92023-03(DRP)). | |||
The inspectors _ reviewed the activities associated to the September 6, | |||
1992, Unit 1 Steam Generator chemistry variation (chloride excursion). | |||
The 10 CFR 50.59 Safety Evaluation of Unit 1 chloride excursion | |||
concludes there is no unreviewed safety question. Inspection Report 50- | |||
456/92020; 50-457/92020, details the chloride excursion experienced by | |||
Unit.1 during its shutdown to commence a refueiing outage. As required | |||
by Technical- Specification 3.4.7, a Fafety Evaluation was completed | |||
which addressed the potential effect of this excursion on the Reactor | |||
Coolant System (RCS) austenitic stainless steel, Alloy 600 materials, | |||
and Zircaloy fuel cladding. | |||
The 10 CFR 50.59 safety evaluation was completed by Westinghouse and | |||
concluded there was no unreviewed safety question resulting from the | |||
excursion. The technical basis for this conclusion was: | |||
< a. Austenitic stainless steels are potentially susceptible to | |||
chloride stress corrosion cracking (SCC) in aqueous solutions | |||
under certain conditions. Oxygen is necessary for chloride | |||
cracking in austenitic steels at temperatures below boiling. No | |||
detectable oxygen was reported in the Unit 1 RCS during the | |||
- | |||
excursion. Hydrogen peroxide was not.added to the RCS during the | |||
excursion. Moreover, of the 47-hour duration of the-excursion, | |||
the RCS average temperature was above 150 F-(the acceleration | |||
. temperature for chloride SCC) for only. 35-hours Existing | |||
Westinghouse test data for sensitized 304 stainless steel- | |||
indicates that-the crack initiation time for chloride | |||
concentrations of 390 ppb (the peak concentration seen during the- | |||
excursion)'is on the order of 12-13 months, well beyond the-35- | |||
hour exposure for-Unit 1. The test data was conservatively based | |||
~ | |||
on exposure in a fully aerated-chloride solution of 150 F. As | |||
previously stated, no detectable oxygen was reported-in the RCS | |||
during the excursion. Therefore, under the conditions for Unit-1, | |||
it was judged that the elevated chloride concentration would not | |||
have a negative impact on-the performance of the austenitic | |||
stainless steel present in the RCS. | |||
b. Alloy 600 materials have generally exhibited good resistance to | |||
chloride induced cracking. Existing test data has demonstrated . | |||
8 | |||
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, -- | |||
es re -r - - | |||
,e-- w e- w n -- _ | |||
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4 | |||
that. chloride induced cracking 'is not an issue for Alloy 600 | |||
~ | |||
material exposed to the type of environment found in the Unit 1 - | |||
RCS during the excursion. Alloy 600 c-ring test specimens | |||
stressed to 2/3 of yield strength, have''been exposed to fully- | |||
aerated' chloride solutions with different ciloride t __ concentrations - | |||
(100-1000 ppm) at 150*F for periods of up to 12 months _ with no | |||
-incidents.of cracking. The test data far exceeds the chloride | |||
level and exposure-period that the Alloy 600 materials were | |||
exposed to during the chloride excursion. Therefore, future | |||
performance of Alloy 600 material will not be adversely affected | |||
because of the excursion. | |||
c. Westinghouse Zircaloy-4 material specifications limit the chloride | |||
content to 20 ppm maximum, but certifications for the material | |||
typically report that values are less than 10 ppm. During reactor | |||
operation, the protective oxide that is formed on the Zircaloy-4 | |||
fuel components is considered to contair., due to diffusion- | |||
effects, similar impurity levels to -the base material. Assuming | |||
therefore, that the oxide film on the Unit 1 Zircaloy-4 fuel | |||
components contained approximately 10 ppm chloride prior to the | |||
coolant chloride excursion, and that all of the 390 ppb chloride | |||
from the coolant was absorbed by the oxide 'Im, the total | |||
, resultant chloride content of the oxide fih: would show an | |||
increase of less than 1 ppm. The resulting c.loride content would | |||
still be considerably below the material specification limit of | |||
10 ppm maximum. Additionally, corrosion studies performed on | |||
. Zirconium, where Zirconium at 650 F was placed in water .containing | |||
200 ppm chlorine gas, have shown Zirconium's corrosion resistance | |||
to be much less sensitive to impurities in the water than.to those | |||
in the metal. Thus, no adverse effects on the Zircaloy fuel | |||
cladding, fuel integrity and fuel handling operations are | |||
expected. | |||
6. Onerational Safety Verification (71707) | |||
_ | |||
The inspectors verified that the facility was being operated in | |||
conformance with the licenses and regulatory requirements and that the | |||
licensee's management control system was effectively cerying out its | |||
responsibilities for safe operation. No violatiens or deviations were | |||
, | |||
identified. | |||
* Untested code safety valves. | |||
* Inspectors' concerns. | |||
o inspectors' review of Technical specifications. | |||
* Determination of having proper lift setpoints. | |||
* Unit 1 outage to investigate generator cooling system. | |||
* Westinghouse _ recommendations for generator cooling syst. | |||
* Unit I return to 100% reactor power. | |||
Untested Main Steam Line Code Safety Valves raise questions regarding | |||
the ability of Unit I to proceed with modo change to Mode 3. On | |||
October 29,1992, Un.c 1 entered Mode 3 following completion of its | |||
' refueling outage. During the outage, five Main Steam Line (MSL) Code | |||
Safety valves .were modified or repaired. The Restart Onsite Review | |||
9 | |||
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_ . _ . _ _ _ __ _ _ _ _ __ _ . _ _ . _ _ _ _ _ _ _ _ _ | |||
_ _ _ _ _ - _ _ _ _ . _ _ _ | |||
e | |||
identified the need to test the lift setpoint pressures of these valves | |||
at nominal operating pressure (N0P) and temperature (NOT), while in | |||
Mode 3, to close the Unit 1 outstanding Nuclear Work Requests (NWRs). | |||
In completing the checklist required before entry into Mode 3, the | |||
Unit 1 Supervisor noted that four of the MSL Code Safety valves were | |||
located on the same MSL. Based on the Unit Supervisor's interpretation | |||
of Technical Specification (TS) 3.7.1.1, " Safety Valves," it was | |||
questioned whether the mode change could be made with four untested | |||
(inoperable) code safeties on one main steam line. Table 3.7-1 of the | |||
IS only provides direction for continued power operation witn a maximum | |||
of three inoperable code safety valves. | |||
Through a series of conference calls, the licensee determined "a review | |||
of NWRs associated with these valves provides a reasonable assurance of | |||
croper lift setpoint." Therefore, the mode change could occur since the | |||
valves, although not tested, were not ir. operable. | |||
On October 30, 1992, with Unit 1 at NOP and NOT in Mode 3, the five MSL | |||
Code Safety valves were tested per Braidwood TS Pr edure BwVS 7.1.1-1, | |||
' Main Steam Safety Valves Operability Test." All five valves f ailed the | |||
surveillance, were adjusted and retested satisfactorily. | |||
In reviewing this event, the inspectors raised the following | |||
questions / concerns: | |||
a. Was the interpretation of the TS correct? | |||
Based on the licensee's interpretation, the inspectors then questioned: | |||
b. How was the conclusion that the untested MSL Code Safety valves | |||
had a " reasonable assurance" of having proper lif t setpoints | |||
determined? - | |||
_ | |||
c. Why were the MSL Code Safety valves not considered inoperable and | |||
an operability determination conducted per Braidwood | |||
Administrative Procedure BwAP 330-10, " Operability Determination | |||
of Safety-Related Equipment?" | |||
In reviewing TS 3.7.1.1, che inspectors determined that the provisi'>ns | |||
of TS 3.0.4 applied and the Mode 3 change with four untested /inopr:rable | |||
code safety valves on the same MSL was not prohibited. AltE~igh .he TS | |||
applies to Modes 1, 2, and 3, it is based on the operability of the MSL | |||
Code Safety valves to ensure that secondary coolant system will be | |||
limited to within 110% (1320 psia) of its design pressure of 1200 psia | |||
during the most sever anticipated system operational trans:ent. The | |||
maximum relieving capacity is associated with a turbine trip from 102% | |||
rater thermal power coincident with an assumed loss of condenser heat | |||
sink (i.e., no steam dumps to the condenser). Therefore, since the | |||
ability to provide relieving capacity is not an issue in Mode 3, the | |||
mode change could be made with all five MSL Code Safety valves, on any | |||
single steam line, inoperable. To allow entry into Mode 3, where the | |||
code safety valves can be tested and set, tha provisions of TS 3.0.4 | |||
(the mode change provision) are applicable to TS 3.7.1.1, | |||
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The determination that the MSL code Safety valves had a reasonable | |||
Assurance of having proper lift setpoints was, in the inspectors | |||
opinion, weak. While a review of the NWRs showed the valves were | |||
assembled correctly, they had not been bench tested. Additionally, | |||
historical data and the engineering judgment of the Technical Staff and | |||
Engineering and Nuclear Construction staff did not support the | |||
conclusion reached. The inspectors questioned how the available | |||
information and engineering judgment was weigFed in reaching a | |||
determination of " reasonable assurance." | |||
Regarding the issue of operability, the inspectors feel there can be no | |||
question that the valves were clearly inoperable and required an | |||
operability determination made per BwAP 330-10. | |||
The inspectors have discussed these questions and concerns with the | |||
licensee and will follow their resolution closely. | |||
Unit I enters planned forced outage to investigate elevated Delta T on ! | |||
the Generator Cooling system. On November 20, 1992, Unit 1 entered Mode | |||
3 to commence a planned forced outage. The outage was required tc | |||
investigate and repair the cause for the elevated temperature | |||
differential (Delta T) of approximately 10 C between the stator-rator | |||
cooling water inlet and outlet temperatures. | |||
Q]' | |||
A review of the refueling outage activities completed on the tator- | |||
rotor cooling svstem did not identify a possible cause of the high ' | |||
Delta T. The licensee performed a fiber-optic inspection of the inlet | |||
manifold, tock flow measurements of individual coils while conducting a | |||
reverse flow flush on the stator, and inspected the coil hoses for | |||
blockage. These efforts failed to identify a root cause for the | |||
condition. | |||
Based on Westinghouse recommendations, the alarm setpoints were raised | |||
to 14 C for high Delta-T and 16oC for high-high Delta-T. Additionally, | |||
Westinghouse performed a balance of the #5 turbine generator bearing. | |||
The balancing successfully reduced the vibration form approximately 5.9 | |||
mils to approximately 1.1 mils. All other bearing vibrations are less 1 | |||
than 2.1 mils. The licensee also repaired various steam, water, and oil M | |||
leaks which had developed since the unit was returned to service 3 | |||
following the refueling outage. | |||
On November 24, 1992, the unit entered Mode 1 and is currently at 100% | |||
uactor power. The licensee and the inspectors will continue to monitor | |||
tne cooling system performance, the inspectors will evaluate the | |||
licensee's corrective actions during a subsequent inspection based on | |||
the system's continued performance. | |||
7. Monthly Surveillance Observation G1726) | |||
The inspectors observed several of the surveillance testing required by | |||
technical specifications during the inspection period and verified that | |||
testing was performed in accordance with adequate procedures, that test | |||
instrumentation was calibrated, that results conformed with technical | |||
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_ _ _ _ - _ _ _ _ _ - _ - _ _ - _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ ________ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ | |||
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specifications and procedure requirements _and were reviewed, and that | |||
any deficiencies identified during the testing were properly resolved. | |||
No violations or deviations were identified. | |||
The following surveillance activities were observed and reviewed: , | |||
* Eddy current inspection. | |||
* 174 tubes were plugged. | |||
* 21 tube plugs were replaced. | |||
Braidwood Technical Surveillance 4.5.0-1, Steam Generator Eddy Current | |||
Inspection. During Unit l's ret'ueling outage, a steam generator eddy | |||
current inspection _ was performed on 100% of the tubes in all four steam | |||
generators from the hot leg tube end through the U-bend. In addition, - | |||
50% of the tubes were examined full length from tube end to tube end. A | |||
total of 174 tubes were plugged due to indications at the hot leg | |||
support plates and antivibration bar (AVB) wear. | |||
In addition to the tubes plugged, a total of 21 Inconel 600 mechanical | |||
tube plugs were replaced with Inconel 690 mechanical tube plugs. This | |||
was accomplished in accordance with NRC Bulletin 89-01, " Failure of | |||
Westinghouse SG Tube Mechanical Plugs." | |||
8. Report Review | |||
During the inspection period, the inspector reviewed the licensee's | |||
Monthly Performance Report for October 1992. The inspector confirmed | |||
that the information provided met the requirements of Technical | |||
Specification 6.9.1.8 and Regulatory Guide 1.16. | |||
The inspector also reviewed the licensee's Monthly Plant Status Report | |||
for September 1992. | |||
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No violations or deviations were identified. | |||
9. Exit Interview (307031 | |||
The inspectors met with the licensee representatives-denoted in | |||
Paragraph I during the inspection period and at the conclusion of the | |||
inspection on November 30, 1992. The inspectors summarized the scope | |||
and results of the inspection and discussed the likely content of this | |||
inspection report. -The licensee acknowledged the information and did | |||
not indicate that any of the information disclosed during the inspection | |||
could be considered proprietary in nature. | |||
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}} | }} |
Latest revision as of 23:19, 23 July 2020
ML20126E832 | |
Person / Time | |
---|---|
Site: | Braidwood ![]() |
Issue date: | 12/17/1992 |
From: | Farber M NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
To: | |
Shared Package | |
ML20126E829 | List: |
References | |
50-456-92-23, 50-457-92-23, NUDOCS 9212300007 | |
Download: ML20126E832 (12) | |
See also: IR 05000456/1992023
Text
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U.S. NUCLEAR REGULATORY COMMISSION
REGION III .
Reports No. 50-456/92023(DRP); 50-457/92023(DRP)
Docket Nos. 50-456; 50-457 Licenses No. N?F-72; NPF-77
Licensee: Commonwealth Edison Company
Opus West III
1400 Opus Place
Downers Grove, IL 60515
Facility Name: Braidwood Station, Units 1 and 2
Inspection At: Braidwood site, Braidwood, Illir.ais
Inspection Conducted: October'13 through November 30, 1992
Inspectors: S. G. Du Pont
J. R. Roton
G. M. Hausman
Approved By: M.Farbh,C M 7!92-
Reacto/ProjectsSectionlA Date
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Inspection Summary-
s
Inspection from October 13 throuch November 30. 1992 (Reports No. 50-
456/92023(DRP): 50-457/92023(DRP))
Areas Inspected: Routine,. unannounced safety inspection by the resident and
regional inspectors of licensee action on previously identified items;
licensee event report review; outages; radiation protection; operational
safety verification; mont!.ly surveillance observation; and report review.
Results: Three violations were identif':ed in one of the six areas inspected.
In the remaining areas, no violations were identified.
The. following is a summary of the licensee's performance during this
inspection period:
I
L Plant Ooerations
The licensee's performance in this area for this inspection period.was1
good. Shift briefings continued to provide sufficient'information for.
l planned evolutions to be performed during the shift. The. inspectors
have raised several questions involving operability determinations
associated with the Main Steam Line Code Safety Valves.-
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l .9212300007 921218 - '
PDR ADOCK 05000456'-
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Radiolooical Controls-
Three violations were issued due to the licensee's failure to adequately
control the addition of SF -to the steam generators (two violations) and
the failure to adhere to the posting requirements' of Radiologically
Controlled' Areas'. Additionally, the report discusses activities
associated with safety evaluations for the SF, and a chloride excursion.
One was an example of good efforts producing a detailed evaluation and
-the other was an example of a failure to perform an evaluation.
Safety Assessment /0uality Verification
The one LER reviewed during this inspection period appears to have
appropriate corrective actions to preclude similar events. The
licensee's evaluation of the Unit I chloride excursion is a good example
of a detailad and comprehensive safety assessment. However, the failure
to conduct a similarly comprehensive evaluation for the sulfur
hexafluoride addition indicates that the sensitivity to and
understanding of the need for safety assessments is not uniform
throughout the licensee's organization.
Enaineerina and Technical Support
Due to the inspectors limited review in this area, the licensee's
performance was not assessed for this inspection period.
Maintenance and' Surveillance
The licensee's performance in maintenance and surveillance during this
inspection period was good.
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DETAILS-
1 -. Persons Contacted ,
Commonwealth Edison Comnany (CECO)
- K. L. Kofton,-Station Manager ,
G. R. Masters, Project Manager
G. E. Groth, Production Superintendent
D. E. O'Brien, Technical' Superintendent
D. E. Cooper, Assistant Superintendent - Operations '
R. J. Legner, Services Director
A. O Antonio, Nuclear Quality Program Superintendent .
- R. Byers, Assistant Superintendent Work __ Planning
G. Vanderheyden,. Technical Staff Supervisor
S. Roth, Security Administrator
K. G. Bartes, Nuclear Safety Supervisor '
A. Haeger, Regulatory Assurance Supervisor
- J. Lewand, Regulatory Assurance
S. Hunsader, EQ Supervisor Design Support - Nuclear Engineering
K. C. Radke, Technical Staff System Engineer
- Denotes those attending the exit interview conducted on November 30,
1992.
The inspectors also interviewed several other licensee employees.
2. Licensee Action on Previously identified Items (92701. 927021
a. Ol en item
LClosed 50-456/92017-02(DRP): 50-48i7 /92017-02(DRPl: Failure to
Mlow Posting Requirements of a Radiologically Controlled Area.
Inspection Report 92017 details the f ailure of a Radiological
4 Protection Technician (RPT) to adhere to the posting requirement
to conduct 'a whole body frisk prior to exiting _ a radiologically
controlled area (RCA). In their followup review, the inspectors
'
discovered that-two weeks prior to this incident, a RPT had failed
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to verify the decontamination of the 1A letdown heat exchanger
room before removing the posting. As a result, the RPT and one
- Electrical Maintenance Department person were contaminated when
they entered the room to replace light bulbs. At:ditionally, there -
has been one other incident since the open item was identified. In
this incident, _two Mechanical Maintenance Department personn_el
failed to adhere to the posted requirements-for entry into a RCA
and were subsequently contaminated. These failures to adhere to
the posted requirements for conducting work within a RCA are
,
violations of Braidwood Technical Specification 6.11, " Radiation
Protection Program," as detailed in Braidwood Radiation Protection
Procedure 1110-3, " Radiological Postings, Labels, and Controls,"
(50-456/92023-01(DRP); 50-457/92023-01(DRP)).
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(Closed) ODen Item (456/88010-Ol(DRS):457/880ll-01(DRS)):
Adequacy of fire protection for several unprotected structural
steel columns and auxiliary steel attachments. The columns were
located in the fuel building between V and W at coordinates 17,
18, and 19, and the attachments were in the auxiliary building on
column P-21 at elevation 401'-0". The~ licensee was to provide the l
methodology used for selection and_ identification of the fire 1
protected structural steel components and the technical 1
justification that column P-21 met the specified fire rating. l
For the culumns, the licensee stated that because of the low fire
loading and the large open volume of the area, a credible fire
would pose no hazard to the structural steel columns, therefore,
fire proofing of the columns was not required. Justification was
provided in Sargent & Lundy Engineers letter dated May 20, 1988.
The columns support the slab at elevation 451' 0", a portion of
which carries a fire rating. The calculated fire loading for the
area, which includes an allowance for transient combustibles, is
5000 Stu/ft' (Fire Protection Report, Subsection 2.3.12.1). This
equates to a fire severity of under four minutes duration (NFPA-
Fire Protection Handbook, Chapter 9, Section 7). Therefore, a
credible fire would pose no hazard to the structural steel
columns.
For the auxiliary steel attachments, the licensee stated that- the
additional heat transfer into the fire protected column from the
unprotected auxiliary steel attachments did not degrade the fire
rating for column P-21 below the specified three hour rating.
Justification was supplied in Sargent & Lundy Engineers letter
dated May 20, 1988. The column was protected by a fire-proof
material, Pyrocrete 102 (7/8" thick), in accordance with
applicable installation drawings, which designated a three hour
fire rating according to Underwriters Laboratory (UL) Detail
X-719. The UL rating was based on tests conducted on a W10x49
column. P-21 was a W14x342 column, which had a cross section
seven times as massive as the VL tested column. The American Iron
and Steel Institute (AISI) had performed extensive research and
tests on a wide range of column sizes including sections which
were more massive than the UL tested W10x49 column. These tests
were summarized in AISI publication " Design Fire Protection for
i Steel Columns," Third Edition, March 1980, which indicated that
L the effective fire rating of the W14x342 column was more than
t twice that for the W10x49 column. Therefore, ample margin was
L provided to compensate for the additional heat input from a
i potential fire due to the unprotected auxiliary steel attachments.
Based upon the above, the inspectors concluded that the
methodology and technical justification provided were acceptable '
and the inspectors had no further concerns. This item is closed.
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b. Unresolved Items
(Closed) Unresolved Item (456/90Q19-02(DRSl: Physical separation
between fuel oil overflow, supply, and vent lines associated with
the opposite train emergency diesels. The licensee conducted a
detailed review'and analysis ~that determined the installed piping
arrangement, although, not consistent with the configuration
described in the fire hazard analysis-(FHA), posed no immediate
operational concerns for the opposite train diesel.. The FHA
stated that all equipment, cables, and piping in the diesel
generator room, the diesel oil day tank room, and the diesel oil-
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storage tank room would be associated with only one ESF division.
However, physical inspections revealed that some of the diesel oil
piping on each train was routed through the three rooms of the
opposite train.
.
A detailed review of the Byron /Braidwood Fire Protection Report
(FPR) was made to identify all other inconsistences related to
physical separation of safety related equipment, cables, and
piping. This consisted of a zone-by-zone review of the FHA
(Section 2.3 of the FPR), a review of the safe shutdown analysis
(Section 2.4 of the FPR), and a review of. FPR Section 3.0, which
addressed conformance to the Standard Review Plan. No other
inconsistences from the FPR were identified. Minor changes were
completed to address the inconsistences as described in Sargent &
Lundy Engineers Letter dated November 29, 1990, and'Transmittcl
DIT-88-EXT-0124, " Assessment of Diesel Oil Piping Routed in
Opposite Train Diesel Generator Rooms" dated February 25, 1992.
The inspectors reviewed the licensee's detailed analysis and
discussed the results with NRR. The discussion concluded that the
licensee's analysis was acceptable and that Appendix R concerns
were adequately addressed, since offsite power would be available
and the diesel oil piping would remain intact during a postulated
fire. . Corrective actions taken by the licensee indicated prompt.
actions were performed including an expanded scope of review and
analysis which included the Byron Station, as-well-as, evaluating
the probability of missile affects on the opposite train diesel
piping. The inspectors noted, however, that during the
performance of two minor changes numerous field problem. reports
(FPRs) were generated relating to interference and clearance
. problems indicating that planning was not effective. The
inspectors had no further concerns and considered this item
closed.
c. Violation
(Closed) 50-456/92017-03(DRP): 50-457/92017-03(DRP): Technical
Specification 6.8.1 was violated when the licensee failed to-
convert Nuclear Work Requests to Temporary Alterations in
accordance with Braidwood Administrative Procedure 2321-18. 'The
licensee's response to-this violation was prompt and thorough.
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Actions taken appear adequate to preclude recurrence of this or
similar events. This item is closed.
One violation was identified.
3. Licensee Event Report (LER) Review (92700)
LER: were reviewed and closed based on the following criteria:
- Reportability requirements were met.
- Immediate corrective actions were accomplished.
- Corrective actions to prevent recurrence has been or will be
initiated per technical specifications.
No violations or deviations were identified.
(Closed) 457/01006: Reactor Trip Due to Valve Hispositioning. This
event is discussed, in detail, in Inspection Report 50-456/92020; 50-
457/92020, Paragraph 4. In addition to those corrective actions
previously identified, the inspector also notes the involvement of GN
Venture contractor personnel in the assessment of root cause
determination and corrective action. This item is closed.
4. Outanes (86700)
No violations or deviations were identified.
. On November 3,1992, Unit 1 Main Generator was synchronized to the grid,
ending the licensee's planned 66-day refueling outage seven days ahead
of schedule. In addition to completing ahead of schedule, the 185.0 Rem
of exposure and 88 personnel contaminations were well under the ALARA
goals of 208.5 Rem and 126 personnel contaminations established for the
outage. From a budgetary standpoint, initial estimates show the outage
to be $1.8 million under the approved budget. The outage was -
accomplished without major incidents and difficulties. The lessons-
learned must be carried forward into the refueling outage for Unit 2,
which begins in March 1993.
M 5. Radiation Protection (937011
Two violations were identified pertaining to radiological work
practices, performing safety analysis, and preparing written procedures
for chemistry related evolutions.
,
Addition of sulfur hexafluoride (SF.).
- Evaluation of planning and implementation (SF.).
- Inspectors conclusion (SF ).
- Apparent violations (SF.) .
- November 6,1992, chloride excursion.
- Chloride excursion safety evaluation.
Addition of sulfur hexafluoride (SF.) causes unexpected variations in
steam generator chemistry. On November 6, 1992 the Braidwood Chemistry
Department inadvertently caused a chemical variation on the Unit 1 steam
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generators. Technicians injected a large amount, Labout 15 standard
cubic; feet, of sulfur hexafluoride (SF.) gas into the condensate system.
The injection of SF, was part of troubleshooting efforts on Unit 2--steam '!
generators. _ Probable leakage in the Unit' 2. steam generators resulted in .i
unexpected chemistry leve'is. Since the condensate system contains
several connections between the units, the licensee ;uspected that these i
connections were.the leakage source.
'
Shortly after injecting SF., the technicians noticed unexpect' ed
responses in the Unit 1 steam generators' chemistry. A chemistry sample
confirmed that a large variation occurred. This required entry into the
action level of the station chemistry. procedures. The sample displayed
elevated cation conductivity, phosphates, fluorine, and sulfonates. -The
technicians determined that the SF, gas unexpectedly broke down into
sulfonates and fluorine.
The inspectors evaluated the planning and implementation of _the
troubleshooting efforts involving injection of SF ._ Previously, SF,
injection was used to find leaks in the condenser water boxes. The
success of this usage influenced the chemistry department to use SF, in
the condensate.
-
However, they did not evaluate the possibility of SF, going in the steam
generators. They also did not evaluate the effects of the -steam
generator water chemistry on SF, . S F, gas is only slightly soluble in
water and soluble in alkaline solutions. Since water chemistry is-
alkaline, SF, broke down into fluorine and sulfonates. This condition
is not desirable since fluorine is corrosive to the heat transfer
surfaces and sulfonates is basic. The technicians injected SF, into
the condensate system without considering th'ese effects. They also
injected the SF, without a procedure.
The inspectors concluded that the chemistry department did not follow
several procedures and requirements before and during the injection of
the SF. gas. 10 CFR 50.59, " Changes, Tests,-and Experiments," requires .'
performance of an evaluation of tests or experiments not described in
the safety analysis report. The regulation also requires the . evaluation-
for changes in the facility. _This evaluation is to determine the
possibility of an u_nreviewed safety question or a change in the
1 technical specification.
The Braidwood safety a'nalysis report describes the methods for
naintaining water chemistry in the steam generators. The addition of
ammonia in the form of ammonium hydroxide, or an equivalent amine, and
hydrozine to the condensate is in Section 10.3.5.1. The addition of SF,
gas to_the condensate is not in-the safety analysis report.
TD regulation 10 CFR 50, Appendix B, Criterion V, " Instructions,
< Procedures, and Drawings,' requires that activities affecting quality
shall be done by documented procedures. These. procedures shall be of a
type appropriate to the circumstances of the activity and followed.
Additions *,f chemicals to the condensate affects the steam generator and
can affect the quality _of the heat transfer surface.
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The Braidwood Technical Specification, Section 6, " Administrative
Controls'," requires-the_ establishing _of procedures for activities
prescribed in Regulatory Guide 1.33,. Quality Assdrance Program
Requirements. The regulatory _ guide requires the establishing of
procedures for controlling water quality. These procedures _will contain
limits for concentrations of agents that are corrosive to heat transfer
surfaces. The addition of SF. into the ccndensate was performed without--
establishing a procedure.
The inspectors concluded that the following activities were violations
of NRC requirements. The injection of SF, without a procedure was an
apparent violation of Technical Specifications (50-456/92023-02(DRP);
50-457/92023-02(DRP)). The failure to perform an evaluation for
unreviewed safety condition is an apparent violation of 10 CFR 50.59
(50-456/92023-03(DRP); 50-457/92023-03(DRP)).
The inspectors _ reviewed the activities associated to the September 6,
1992, Unit 1 Steam Generator chemistry variation (chloride excursion).
The 10 CFR 50.59 Safety Evaluation of Unit 1 chloride excursion
concludes there is no unreviewed safety question. Inspection Report 50-
456/92020; 50-457/92020, details the chloride excursion experienced by
Unit.1 during its shutdown to commence a refueiing outage. As required
by Technical- Specification 3.4.7, a Fafety Evaluation was completed
which addressed the potential effect of this excursion on the Reactor
Coolant System (RCS) austenitic stainless steel, Alloy 600 materials,
and Zircaloy fuel cladding.
The 10 CFR 50.59 safety evaluation was completed by Westinghouse and
concluded there was no unreviewed safety question resulting from the
excursion. The technical basis for this conclusion was:
< a. Austenitic stainless steels are potentially susceptible to
chloride stress corrosion cracking (SCC) in aqueous solutions
under certain conditions. Oxygen is necessary for chloride
cracking in austenitic steels at temperatures below boiling. No
detectable oxygen was reported in the Unit 1 RCS during the
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excursion. Hydrogen peroxide was not.added to the RCS during the
excursion. Moreover, of the 47-hour duration of the-excursion,
the RCS average temperature was above 150 F-(the acceleration
. temperature for chloride SCC) for only. 35-hours Existing
Westinghouse test data for sensitized 304 stainless steel-
indicates that-the crack initiation time for chloride
concentrations of 390 ppb (the peak concentration seen during the-
excursion)'is on the order of 12-13 months, well beyond the-35-
hour exposure for-Unit 1. The test data was conservatively based
~
on exposure in a fully aerated-chloride solution of 150 F. As
previously stated, no detectable oxygen was reported-in the RCS
during the excursion. Therefore, under the conditions for Unit-1,
it was judged that the elevated chloride concentration would not
have a negative impact on-the performance of the austenitic
stainless steel present in the RCS.
b. Alloy 600 materials have generally exhibited good resistance to
chloride induced cracking. Existing test data has demonstrated .
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that. chloride induced cracking 'is not an issue for Alloy 600
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material exposed to the type of environment found in the Unit 1 -
RCS during the excursion. Alloy 600 c-ring test specimens
stressed to 2/3 of yield strength, havebeen exposed to fully-
aerated' chloride solutions with different ciloride t __ concentrations -
(100-1000 ppm) at 150*F for periods of up to 12 months _ with no
-incidents.of cracking. The test data far exceeds the chloride
level and exposure-period that the Alloy 600 materials were
exposed to during the chloride excursion. Therefore, future
performance of Alloy 600 material will not be adversely affected
because of the excursion.
c. Westinghouse Zircaloy-4 material specifications limit the chloride
content to 20 ppm maximum, but certifications for the material
typically report that values are less than 10 ppm. During reactor
operation, the protective oxide that is formed on the Zircaloy-4
fuel components is considered to contair., due to diffusion-
effects, similar impurity levels to -the base material. Assuming
therefore, that the oxide film on the Unit 1 Zircaloy-4 fuel
components contained approximately 10 ppm chloride prior to the
coolant chloride excursion, and that all of the 390 ppb chloride
from the coolant was absorbed by the oxide 'Im, the total
, resultant chloride content of the oxide fih: would show an
increase of less than 1 ppm. The resulting c.loride content would
still be considerably below the material specification limit of
10 ppm maximum. Additionally, corrosion studies performed on
. Zirconium, where Zirconium at 650 F was placed in water .containing
200 ppm chlorine gas, have shown Zirconium's corrosion resistance
to be much less sensitive to impurities in the water than.to those
in the metal. Thus, no adverse effects on the Zircaloy fuel
cladding, fuel integrity and fuel handling operations are
expected.
6. Onerational Safety Verification (71707)
_
The inspectors verified that the facility was being operated in
conformance with the licenses and regulatory requirements and that the
licensee's management control system was effectively cerying out its
responsibilities for safe operation. No violatiens or deviations were
,
identified.
- Untested code safety valves.
- Inspectors' concerns.
o inspectors' review of Technical specifications.
- Determination of having proper lift setpoints.
- Unit 1 outage to investigate generator cooling system.
- Westinghouse _ recommendations for generator cooling syst.
- Unit I return to 100% reactor power.
Untested Main Steam Line Code Safety Valves raise questions regarding
the ability of Unit I to proceed with modo change to Mode 3. On
October 29,1992, Un.c 1 entered Mode 3 following completion of its
' refueling outage. During the outage, five Main Steam Line (MSL) Code
Safety valves .were modified or repaired. The Restart Onsite Review
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identified the need to test the lift setpoint pressures of these valves
at nominal operating pressure (N0P) and temperature (NOT), while in
Mode 3, to close the Unit 1 outstanding Nuclear Work Requests (NWRs).
In completing the checklist required before entry into Mode 3, the
Unit 1 Supervisor noted that four of the MSL Code Safety valves were
located on the same MSL. Based on the Unit Supervisor's interpretation
of Technical Specification (TS) 3.7.1.1, " Safety Valves," it was
questioned whether the mode change could be made with four untested
(inoperable) code safeties on one main steam line. Table 3.7-1 of the
IS only provides direction for continued power operation witn a maximum
of three inoperable code safety valves.
Through a series of conference calls, the licensee determined "a review
of NWRs associated with these valves provides a reasonable assurance of
croper lift setpoint." Therefore, the mode change could occur since the
valves, although not tested, were not ir. operable.
On October 30, 1992, with Unit 1 at NOP and NOT in Mode 3, the five MSL
Code Safety valves were tested per Braidwood TS Pr edure BwVS 7.1.1-1,
' Main Steam Safety Valves Operability Test." All five valves f ailed the
surveillance, were adjusted and retested satisfactorily.
In reviewing this event, the inspectors raised the following
questions / concerns:
a. Was the interpretation of the TS correct?
Based on the licensee's interpretation, the inspectors then questioned:
b. How was the conclusion that the untested MSL Code Safety valves
had a " reasonable assurance" of having proper lif t setpoints
determined? -
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c. Why were the MSL Code Safety valves not considered inoperable and
an operability determination conducted per Braidwood
Administrative Procedure BwAP 330-10, " Operability Determination
of Safety-Related Equipment?"
In reviewing TS 3.7.1.1, che inspectors determined that the provisi'>ns
of TS 3.0.4 applied and the Mode 3 change with four untested /inopr:rable
code safety valves on the same MSL was not prohibited. AltE~igh .he TS
applies to Modes 1, 2, and 3, it is based on the operability of the MSL
Code Safety valves to ensure that secondary coolant system will be
limited to within 110% (1320 psia) of its design pressure of 1200 psia
during the most sever anticipated system operational trans:ent. The
maximum relieving capacity is associated with a turbine trip from 102%
rater thermal power coincident with an assumed loss of condenser heat
sink (i.e., no steam dumps to the condenser). Therefore, since the
ability to provide relieving capacity is not an issue in Mode 3, the
mode change could be made with all five MSL Code Safety valves, on any
single steam line, inoperable. To allow entry into Mode 3, where the
code safety valves can be tested and set, tha provisions of TS 3.0.4
(the mode change provision) are applicable to TS 3.7.1.1,
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The determination that the MSL code Safety valves had a reasonable
Assurance of having proper lift setpoints was, in the inspectors
opinion, weak. While a review of the NWRs showed the valves were
assembled correctly, they had not been bench tested. Additionally,
historical data and the engineering judgment of the Technical Staff and
Engineering and Nuclear Construction staff did not support the
conclusion reached. The inspectors questioned how the available
information and engineering judgment was weigFed in reaching a
determination of " reasonable assurance."
Regarding the issue of operability, the inspectors feel there can be no
question that the valves were clearly inoperable and required an
operability determination made per BwAP 330-10.
The inspectors have discussed these questions and concerns with the
licensee and will follow their resolution closely.
Unit I enters planned forced outage to investigate elevated Delta T on !
the Generator Cooling system. On November 20, 1992, Unit 1 entered Mode
3 to commence a planned forced outage. The outage was required tc
investigate and repair the cause for the elevated temperature
differential (Delta T) of approximately 10 C between the stator-rator
cooling water inlet and outlet temperatures.
Q]'
A review of the refueling outage activities completed on the tator-
rotor cooling svstem did not identify a possible cause of the high '
Delta T. The licensee performed a fiber-optic inspection of the inlet
manifold, tock flow measurements of individual coils while conducting a
reverse flow flush on the stator, and inspected the coil hoses for
blockage. These efforts failed to identify a root cause for the
condition.
Based on Westinghouse recommendations, the alarm setpoints were raised
to 14 C for high Delta-T and 16oC for high-high Delta-T. Additionally,
Westinghouse performed a balance of the #5 turbine generator bearing.
The balancing successfully reduced the vibration form approximately 5.9
mils to approximately 1.1 mils. All other bearing vibrations are less 1
than 2.1 mils. The licensee also repaired various steam, water, and oil M
leaks which had developed since the unit was returned to service 3
following the refueling outage.
On November 24, 1992, the unit entered Mode 1 and is currently at 100%
uactor power. The licensee and the inspectors will continue to monitor
tne cooling system performance, the inspectors will evaluate the
licensee's corrective actions during a subsequent inspection based on
the system's continued performance.
7. Monthly Surveillance Observation G1726)
The inspectors observed several of the surveillance testing required by
technical specifications during the inspection period and verified that
testing was performed in accordance with adequate procedures, that test
instrumentation was calibrated, that results conformed with technical
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specifications and procedure requirements _and were reviewed, and that
any deficiencies identified during the testing were properly resolved.
No violations or deviations were identified.
The following surveillance activities were observed and reviewed: ,
- Eddy current inspection.
- 174 tubes were plugged.
- 21 tube plugs were replaced.
Braidwood Technical Surveillance 4.5.0-1, Steam Generator Eddy Current
Inspection. During Unit l's ret'ueling outage, a steam generator eddy
current inspection _ was performed on 100% of the tubes in all four steam
generators from the hot leg tube end through the U-bend. In addition, -
50% of the tubes were examined full length from tube end to tube end. A
total of 174 tubes were plugged due to indications at the hot leg
support plates and antivibration bar (AVB) wear.
In addition to the tubes plugged, a total of 21 Inconel 600 mechanical
tube plugs were replaced with Inconel 690 mechanical tube plugs. This
was accomplished in accordance with NRC Bulletin 89-01, " Failure of
Westinghouse SG Tube Mechanical Plugs."
8. Report Review
During the inspection period, the inspector reviewed the licensee's
Monthly Performance Report for October 1992. The inspector confirmed
that the information provided met the requirements of Technical
Specification 6.9.1.8 and Regulatory Guide 1.16.
The inspector also reviewed the licensee's Monthly Plant Status Report
for September 1992.
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No violations or deviations were identified.
9. Exit Interview (307031
The inspectors met with the licensee representatives-denoted in
Paragraph I during the inspection period and at the conclusion of the
inspection on November 30, 1992. The inspectors summarized the scope
and results of the inspection and discussed the likely content of this
inspection report. -The licensee acknowledged the information and did
not indicate that any of the information disclosed during the inspection
could be considered proprietary in nature.
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