IR 05000263/1977003
| ML20024G075 | |
| Person / Time | |
|---|---|
| Site: | Monticello  |
| Issue date: | 05/20/1977 |
| From: | Creswell J, Little W, Phillip G NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML20024G067 | List: |
| References | |
| 50-263-77-03, 50-263-77-3, NUDOCS 9102070509 | |
| Download: ML20024G075 (15) | |
Text
{{#Wiki_filter:.. _ .. I .. .. . . , % '(' 0FFICE OF INSPECTION AND EN1'ORCDIENT bNITED STATES NUCLEAR REGULATORY COMMISSION .
REGION III
Report of Operations Inspection , IE Inspection Report No. 050-263/77-03 , Licensce: Northern States Power Company 414 Nicollet Hall Minneapolis, MN 55401 Monticello Nuclear Generating Plant License No. DPR-22 Monticello, MN Category: C , Type of Licensee: BWR' 9E 1670 }Nt Type of Inspection: Special Dates of Inspection: February 25 and March 17 and 18, 1977 N (ic....-e fb q Principal Inspector: 'J.
S. Creswell E MY
/ , [,f date signed Accompanying Inspector: G. A. Phillip [')8' 77 date signed Other Accompanying Personnel: W. S. Little Reviewed By: f f /< / Nuclear Support Section 'date' signed I ' . 9102070509 770616 PDR ADOCK 05000263 O PDR .. ... . !
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SUMMARY OF FINDINGS Inspection Summary Inspections were conducted on February 25 and March 17 and 18, 1977, (77-03): regarding the reactor scram associated with a short period which occurred February 23, 1977, at the Monticello Nuclear Station.
The inspections were held in the corporate offices of Northern Sta',es Power Company located at Minneapolis, Hjnnesotc. and at the Montice)1o Noelcar Plant.
A management conference was held Aptil 12 at the corporate offices.
Two items of noncompliance were identified assoc 1-ated with the failure to follow technical specification procedure requirements.
Enforcement Items Infractions A.
Technical Specification 6.5.A.1 requires that normal startup, operation, and shutdown of the reactor be conducted in accordance with detailed written procedures.
1.
Procedure C.1, Startup, in Part B, Step 3, and ref erenced Part A, Section 2, Step 4.c. requires that the licensee " observe the period meter and govern control rod withdrawal to avoid a ( reactor period shorter than 30 seconds."
Contrary to the above, on February 23, 1977, during a reactor startup, control rods were withdrawn on two separate occasions when there were indications that withdrawing the rod.= would produce periods less than 30 seconds.
Both actiorm resulted in periods less than 30 seconds.
Preceeding these rod with-drawals was a rod withdrawal which resulted in an indicated period of 10 seconds.
The subsequent rod withdrawals were either made with the same rod over the same notch positions, or on a symmetric control rod where response should liave been similar.
(Paragraph 2, Report Details) 2.
Procedure C.1 states in Part D, Section 1 that certain require-ments must be observed prior to startup after a scram.
One of these requirementa is found in Section 3.a and states, "The cause of the scram must be determined and the condition which caused the scram must be corrected."
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' . , . . (~ Contrary to the above, following a scram which occurred at approximately 5:00 p.m. a startup was initiated in that con-trol rod withdrawal began at approximately 7:15 p.m.
on February 23, 1977, without the condition which caused the scram being corrected.
(Paragraphs 2 and 3.f. Report Details) B.
Technical Specification 6.5.A.3 states, " Detailed written pro-cedures, including the applicable check-off and instructions shall be prepared and followed covering. . actions to be . taken to correct abnormal reactivity changes including follow-up actions required after pro 6ective system actions have been { initiated."
l l Contrary to the above, a detailsd ritten procedure covering actions to be taken to correct abnormal reactivity changes had not been prepared.
(Paragraph 3.j., Report Details) Licensee Action on previously Identified Enforcement Items None inspected.
Other Significant Findings A.
Systems and Components None.
( B.
Facility items (Plans and Procedures) None.
C.
Managerial items None.
D.
Deviations None identified.
E.
Status of Previously Reported Unresolved Items None.
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. j -. , _ ... _,.. _ _ . _ _ - _. .. _ .-- _. _ _ _ _ _. - _ _ _ _.. L l -.. ., . . , k . { Management Interviews and Conference Two exit interviews and a corporate management conference were held with licensee representatives to discuss the results of the inspection.
A.
February 25, 1977 Exit Interview (Corporate Office, Minneapolis, Minnesota) 1.
Attendees
Northern States Power company ) ' G. Neils, General Superintendent of Nuclear Power Plants D. Nevinski, Nuclear Engineer R. Anderson, Superintendent of Core Analysis
M. Voth, Nuclear Safety and Technical Services Engineer Nuc1 car Regulatory Commission J. Creswell, Reactor Inspector 2.
Topics Discussed a.
The inspector noted that before the short period scram had occurred, that a symmetric rod had been pulled and that a period in the range.of 7-10 seconds was observed.
( lie asked the licensee why he had not stopped further rod withdrawals until the cause of the short period l had been determined.
The licensee stated that they wanted to determine if the effect was a local phenomenon, b.
The inspector asked the licensee what criteria had been developed af ter the short period (approximately 1 second) incident to prevent recurrence of short periods.
The licensee stated that all necessary plant personnel had been advised of the problem and that the reactor coolant temperature was lowered approximately 100 F so that criticality would be achieved earlier in the rod withdrawal sequence.
The inspector asked if they had I - 4-r .
, . ... ., .. ,g _._- - . - - _ - , _,. t .. .. . % '(~ dition such that they know they did not have high analyzed rod worths based on the lower temperature con- , worth notchen.
The licensee related that the lower temperature condition had not been analyzed, but that based on their judgement it was safe to restart.
(paragraphs 2 and 4. Report Details) B.
March 18, 1977 Ex). Interview (Monticello Nucicar plant) , l , 1.
Attendees j Northern States Power Company G. Neils, Cencral Superintendent of Nuc1 car Power Plants L. Eliason, Monticello Plant bbnager . H. Clarity, Superintendent Plant Engineering and Radiation Protection ' D. Nevinski, Nuclear Engineer ' Nuclear Regulatory Commission W. Little, Chief Nuc1 car Support Section J. Creswell, Reactor Inspector G. Phillip, Special Investigator 2.
Topics' Discussed -
The inspector reviewed the results of the inspection with the licensee.
Areas discussed were operator and management i actions associated with the short period trip incident.
l The first area of concern was explained to be licensee judgement in continuing rod withdrawals when the operators were experiencing unexpected reactor behavior including a , l period of approximately 10 seconds.
It was noted that the I actions taken were contrary to the applicable operating pro- ! cedure which requires that periods less than 30 seconds be avoided.
The licensee was told that not following the pro-cedure constituted an item of noncompliance.
(Paragraphs 2 and 3, Report Details) l The second area of concern discussed relatc3 to not taking corrective ac.tions prior to restart after the scram.
(Paragraphs 2 and 3, Report Details) l l l , ( -5-t I
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_Ayril 12, 1977 Manntement Conference 1.
Attendees Northern States Power Company L. Wachter, Vice President, Power Production and System Operation D. Gilberts, General Manager of Power Production G. Neils, General Superintendent of Nuclear Power Plants , L. Eliason, Monticello Plant Manager M. Clarity, Superintendent Plant Engineering and Radiation Protection D. Nevinski, Nuclear Engineer Nuclear Reuulatory Commission J. Reppler, Region III Director G. Fiorelli, Chief, Reactor Operations and Nuclear Support 11 ranch W. Little, Chief, Nuclear Support Section R. Warnick, Chief, Reactor Projects Section 2 J. Creswell, Reactor lospector 2.
Topics Discussed ( On April 12, 1977, inembers of the Region III staf f met with Northern States Power (NSP) personnel at the NSP corporate office.
The purpose of the meeting was to discuss the short period trip abnortaal occurrence which took place on February 25, 1977, at the Monticello Noelcar Plant.
The inspectors related the facts associated with the event as they understood them and asked for licensee comments in those areas in which they disagreed.
' Discussion of the event focused on the causes which involved: a.
Staff communications and understanding of the reactor condition.
b.
Management rationale Icading to the attempted reactor startup and decision for reactor startup following the scram.
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-{ A sequence of events was presented by the inspectora and each item of the sequence and apparent item of noncompliance were ' discussed in detall.
After the presentation the licensee stated that af ter the f act some of the judgements made by the
operators and other piant personnel were not prudent.
Some of the corrective actions discussed by NSp were: a.
More detailed analysis of rod worths will be performed before rod sequences are impicmented in order to pro- . . vide operating personnel with information about high worth notches.
b.. Operating procedures will be revised to more thoroughly define operator actions to be taken in the event of abnormal reactivity additions which result in short ' periods.
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Persons Contacted L. Wachter, Vice President, Power Production and System Operation D. Gilberts, General Manager of Power Production G. Neils, General Superintendent of Nuclear Power Plants L. Eliason, Monticello Plant Manager .M. Clarity, Superintendent Plant Engineering and Radiation Protection D. Nevinski, Nuc1 car Engineer R. Anderson, Superintendent of Core Analysis M. Voth, Nuclear Safety and Technical Services Engineer L. Seversen, Shift Supervisor W. Boehme, Lead Plant Equipment and Reactor Operator R. Rohland, Plant Equipment and Reactor Operator D. Wagner, Nuclear Engineer 2.
Event Sequence On February 23, 1977, the Monticello Nuclear Plant attempted a reactor startup following a scram from an electrical malfunction.
During the subsequent startup attempts, three short reactor periods (approximately 10 seconds and less) occurred during control rod withdrawals.
The last short period resulted in a high flux reactor ( scram.
The inspectors questioned plant and corporate office personnel concerning those events which led up to the fast periods, the reactor scram, and subsequ-,t evaluation which took place prior to the reactor restart.
Imgbooks, recorder charts, the scram report and the Opcration Committee Minutes were also reviewed.
From this infernation the inspectors put together the following sequence of events: Sequence of Events February 23, 1977, 7:00 a.m.
Reactor scras (electrical problem).
8:00 a.m.
Nuclear Engineer communicated with corporate office physicist regarding nuclear core conditions.
12:00 p.m.
Critical rod predictions provided by Nuclear Engineer.
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preparations for restart completed.
1:30 p.m.
Commenced control rod withdrawal.
3:17 p.m.
Reached predicted core reactivity of 0.99 k; however, no measurable multiplication observed at this time.
At this point, procedure C.1 is reviewed and it states that control rod withdrawal "should" be changed from continuous to notch.
' '4:00 p.m.
Shift Supervisor authorized return to continuous control rod withdrawal because of no observabic multiplication.
4:35 p.m.
Reached predicted critical configuration.
Still no discernable multiplication.
Based on discussions with corporate core physicist regarding new calcula-tions on negative reactivity override, nuclear engineer predicted new critical rod positions.
The nuclear engineer di' not believe the new prediction.
. 5:01 p.m.
Step 15 was in progress.
Control Rod 06-27 was with-drawn from positon 0 (to be withdrawn continuously to position 10).
A 10 second period was observed > as the rod traveled between position 6 to 8.
At this point, operator reinserted the rod to position 4 to ( terminate period.
At this point in tine, an Sr.0, RO, and nuclear engineer were at the control board.
Operating procedures did not cover abnormal reactivity change conditions.
SRO made decision to withdraw rod 06-27 notchwise.
5:03 p.m.
Withdrew Rod 06-27 from position 4 to position 6.
No response observed.
5:04 p.m.
Withdrew Red 06-27 from position 6 to position 8 with resultant 10 second period.
This constitutes non-compliance with Technical Specification 6.5.A.1 and referenced Procedure C.1 which requires steps to be taken to avoid periods of less than 30 seconds.
Rod was inserted to position 6 to terminate period.
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5:06 p.m.
Commenced notch withdrawal of control rod 46-27.
5:12 p.m.
In withdrawing rod 46-27 from position 4 to 6, an approximate 600 second period was achieved.
(600 second value obtained from subsequent analysis of . the SRM chart) 5:15 p.m.
Rod 46-27 was withdrawn from position 6 to 8.
A fast period (N 1 sec) was experienced and the reac*,r scrammed on high flux on the IRM's.
(Repeat of the item of noncompliance.)
5:15 p.m. - 7:00 p.m.
Plant management and corporate office informed of events.
The corporate physicist informed the plant , ' that the two control rods involved had been recognized . as being maximum worth rods and that new calculations would be initiated to determine the maximum worth of control rods as it relates to the rod drop accident.
No hold was put on startup by management pending completion of these calculations.
An entry was made in the Safety Review Follovup Record, "No ( corrective action necessary prior to restart. 7:00 p.m.
All prestartup checklists had been completed.
It should be noted that procedures require the cause of the scram to be determined and the condittoa which caused the scram to be corrected.
7:02 p.m.
Permission given for restart although new calculations of control rod worths had not been provided to plant.
7:15 p.m.
Commenced pulling control rods.
9:30 p.m.
Entry in Operator's Daily log describing the plants position relative to the cause of the scram and the correction of the condition which caused the scram.
(Identified as a late entry in the log.)
10:50 p.m.
Achieved criticalicy.
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Event Discunston l The following information was gathered by the inspectors and is considered pertinent: a.
A trainee was at the reactor console during the approach to critical. He was under the direction of the licensed reactor operator, and the inspectors could find no evidence that his presence had any noticabic effect on the incident.
. b.
During the time immediately prior to the first fast period and up to the reactor scram, the Lead Plcnt Equipment and Reactor Operator (senior operators license) and a Nuclear Engineer (unlicensed) were cognizant of the reactor behavior and the licenced reactor operator's actions. The Shift Supervisor and the Superintendent, Plant Engineering and Radiation Protection were in the control room but were not aware of the reactor behavior and operator actions until the reactor scram.
c.
Prior to 4:35 p.m. the Nacicat Engincer received a call from the corporate office notifying him that their calculations revealed that a negative reactivity of greater than 0.075 Ak due to the effects of Xenon, moderator temperature, burnup, etc., would have to be overcome in going critical.
The curves used by the nucicar engineer to estimate the critical rod configuration had not been calculated for this great a value ( and he had to extrapolate the curve to obtain a critical rod prediction.
Because of this the nuclear engineer questioned the accuracy of the prediction.
d.
Those cognizant of the two fast periods immediately prior to the scram were surprised, but at that time no one thought that the proper action was to stop and evaluate the situation until after the scram occurred.
c.
After the reactor trip at 5:15 p.m. some of the plant staff questior.ed whether the Technical Specif f cation 3.3(a) LCO had been exceeded. This states that "the maximum calculated reactivity that could be added by dropout of any increment of any one control blade will not make the core more than 1.3% Ak supercritical." Calculations were initiated (Section 3, Report Details), but they were not completed until the next day, several hours after the reactor had been restarted.
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A startup was initiated in that control rod withdrawal began at approximately 7:15 p.m. on February 23, 1977, without the condition which caused the scram being corrected.
The high notch worths which were experienced before and during the scram at approximately 5:00 p.m. on the same day, were mainly attributabic to a high moderator temperature and more signifi-cantly, to a high Xenon poisoning condition.
Calculations to determine Xenon decay tines were not completed until approxi-mately 9:30 p.m. at which time control rod withdrawals had progressed to the point that 45 rods had been fully withdrawn and 16 more rods had been withdrawn to position 20.
The failure to take corrective action prior to initiating control rod with-drawal constituten noncompliance with Technical Specification 6.5.A.1 and procedure C.I.
g.
An evaluation was performed, as documented in the Operator's Daily Log at 9:30 p.m., which confirmed that rod coupling had been verified and that a control rod had not dropped. While t,he report discussed xenon decay and temperature drops, no specific corrective action was set forth to prevent recurrence.
h.
The licensee contacted GE-San Jose at an unspecified time the evening of February 23, 1977. The events were described to GE.
According to licensee representatives, GE expressed no alarm over the safety consequences of the events.
( i.
The inspector reviewed the Operations Committee Meeting Minutes dated March 1, 1977 during which the incident was reviewed.
The licensee's final report on the incident, including long term corrective action, had not been completed and, therefore, had not yet been reviewed by the Safety Audit Committee.
j.
The licensee did not have any written, approved procedure to define actions to be taken to correct abnormal reactivity changes, such as those which occurred on February 23, or to deperibe follow-up actions required for such an event.
This is contrary to Technical Specifications 6.5.A.3.
4.
Licensee Analysis and Review Concurrent with the cnsite review of the incident, personnel in the corporate offices were informed of the circumstances associ-ated with the trips and they commenced computer analysis of rod worths for rods 06-27 and 46-27.
The licensee utilizes the Nodal ( - 12 - f ' - _ - _. . -
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The Nodal Code is supplied constants from pDQ-7 llarmony which in turn is supplied four r,roup , cross sections f rom Leopard and 1.aser.
The calculation was per-formed at an caposure of 6789 MWD /MT and assumed that the core had run at 100% power for 50 hours preceding the 7:00'a.m. trip.
' The code models nodal xenon decay and for the rod worth determina-tion calculated xenon concentrations at 10.2 hours after the 7:00 'a.m.
trip.
Since it was known that very near critical conditions (approximately 600 second period) existed at position 6 of rod 46-27, the eigenvalue was normalized to a corrected Kef f of 1.0.
In performing the calculation, the licensee used cross section data generated for 68 F reactor coolant temperatures.
Since the actual reactor coolant temperature was 480 F, a correction for the increased temperature was applied using data accumulated at t.a plant during a startup.
This data correlated the multiplication k cot. " ant of the reactor versus reactor coolant temperature. A delta K (Model) of 0.0039 was also applied to the code calculated eigenvalue to account for analytical-experimental disagreement.
The risults of the calculation for rod 46-27 are shown in Table 1.
As can.be seen by the data for rod 46-27, a rod drop f rom position 00 to position 12 would result in a reactivity addition of: Integral worth at position 12 .0157 delta K Integral worth at critical (position 6) .0062 delta K Supercritical Worth .0095 delta K i This value is below the Technical Specifications limit of 1.3% delta K supercritical.
Resulta of the analysis were supplied to the' plant personnel after the restart at 11:30 p.m.
o As a result of safety review follow-up, the licensee stated that they would review and revise as necessary the rod sequence, pro-cedures for establishing criticality, and procedures for providing operators with information relating to the potential for high worth rods. The licensee concluded that there was no evidence of fuel damage since there was no increase in off-gas activity.
5.
Predicted critical Position Prior to criticality at 5:15 p.m. on February 23 station nucicar engineers calculated the following: . Estimated critical position - step 9, rod 46-35, position 22.
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.. . . .- t Estimated critical position - 1% delta K - step 6, rod 14-39.
( Es*.imated critical position + 1% delta K - e,tep 20, rod 30-27.
As was noted earlier criticality was achieved first on rod 06-27 on position 8 in step 15.
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. TABLE 1 ' Monticello Cycle 5 6789 MWD /STu 10.25 bra after Shutdown Group 8 Step 35 Rod Pattern except 06-27 at 06 Notching out Rod 46-27 Relative , 3D Rod 46-27 A AK Corrected Differential Integral Integral Case Notches _ K Vorth (AK) Worth (AK) Worth
100 1.0332 .0394 .9938
0 .0011
102 1.0343 .0394 .9949 .0011 .052 .0022
04 1.0365 .0394 .9971 .0033 .156 .0029
06 1.0394 .0394 1.0000 .0062 .2% .0047
08 1.0441 .0394 1.0047 .0109 .517 .0018
10 1.0459 .0394 1.0065 .0127 .602 .0030
12 1.0489 .0394 1.0095 .0157 .744 .0025 68-
1.0514 .0394 1.0120 .0182 .863 .0022
24 1.0536 .0394 1.0142 .0204 .967 .0007
32 1.0543 .0394 1.0149 .0211 1.000
0
48 1.0541 .0394 1.0147 .0211 1.000 AK = AK(68 F to 480 F) + AK (Model) 'Iotal Rod Worth * 2.11% AK(68 F to 480 F) +.0433 0 - 10 notches = 1.27% AK(Model) =.0039 0 - 12 notches = 1.57% AK = .0433 +.0039 = .0394 ( -
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