ML20111C213
| ML20111C213 | |
| Person / Time | |
|---|---|
| Site: | Seabrook  |
| Issue date: | 01/07/1985 |
| From: | Devincentis J PUBLIC SERVICE CO. OF NEW HAMPSHIRE |
| To: | Knighton G Office of Nuclear Reactor Regulation |
| References | |
| SBN-746, NUDOCS 8501090218 | |
| Download: ML20111C213 (19) | |
Text
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s SEABROOK STATION Engineering Office January 7, 1985 New Hampshire Yankee Divlelen SBN-746 T.F. Q2.2.2 United States Nuclear Regulatory Commission Washington, D. C. 20555 Attention:
Mr. George W. Knighton, Chief Licensing Branch No. 3 Division of Licensing
References:
(a) USNRC Letter, dated September 29, 1983, " Westinghouse Rod Drop Issue (Board Notification No.83-66A)", from Darrell G. Eisenhut, Director, Division of Licensing, Of fice of Nuclear Reactor Regulation to the Commissioners (b) PSNH Letter, SBN-172, dated July 27, 1981, "10CFR50.55(e)
Interim Report on Rod Drop Analysis", from John DeVincentis to USNRC Region I Office of Inspection and Enforcement (c) PSNH Letter, SBN-143, dated December 16, 1980, "10CFR50.55(e) Interim Report on Rod Drop Analysis", from John DeVincentis to USNRC Region I Office of Inspection and Enforcement (d) PSNH Letter, SBN-131, dated August 21, 1980, "10CFR50.55(e), Interim Report on Rod Drop Analysis", f rom John DeVincentis to USNRC Region 1 Office of Inspection and Enforcement (e) PSNH Letter, SBN-109, dated December 13, 1979, "10CFR50.55(e), Interim Report on Rod Drop Analysis", from John DeVincentis to USNRC Region I Office of Inspection and Enforcement
Subject:
10CFR50.55(e) Final Report on Rod Drop
Dear Sir:
Our Final Report on Rod Drop is attached in the form of annotated FSAR pages. The annotated FSAR pages will be incorporated in the next OL Application amendment. The proposed amendment to FSAR Section 15.4.3 reflects the dropped rod analysis performed by Westinghouse using the methodology described in WCAP-10297-P-A, " Dropped Rod Methodology for Negative Flux Rate Trip Plants".
This analysis has been performed for Seabrook Unit 1, Cycle 1, in accordance with the NRC requirements for a plant and cycle specific analysis, as required by the SER on WCAP-10297.
Verification that the DNB design basis is met for subsequent cycles will be done as part of the standard reload design effort.
O b 8501090218 850107 PDR ADOCK 05000443 S
PDR 1$\\
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P.O. Box 300
- Seobrook.NHO3874 + Telephone (603)474-9521 g (
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'r; United States Nuclear Regulatory Commission Attention:
Mr. George W. Knighton Page 2 In References (b), (c), (d),' and (e), we submitted information to Region I as, required by 10CFR50.55(e).
This letter constitutes the Final Report in accordance with the provisions of 10CFR50.55(e).
Please also note that your approval of the attached amended FSAR Section 15.4.3 will result in removal of the restriction on operations above 90% power which require that the reactor be in manual control or the rods be greater than 215 steps out (Reference Seabrook' Safety Evaluation Report
' +Section 15.4.3; NUREG-0896).
Very truly yours, Q
I John DeVincentis, Director 9.
'(
Engineering and Licensing Attachment.
cc: - Atomic Safety and Licensing Board Service List United States Nuclear Regulatory Commission Region I 631 Park Avenue King of Prussia, PA 19406,
Attention:
Mr. Richard W. Starostecki, Director Division of Project and_ Resident Programs 4
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' William S. Jordan, III Diana Curran Hermon, Weiss & Jordan 20001 S Street N.W.
Brentwood Board of Selectmen Suite 430 RED Dalton Road Washington, D.C.
20009 Brentwood, New Hampshire 03833 Robert G. Perlis Office of the Executive Legal Director Edward F. Meany U.S. Nuclear Regulatory Commission Designated Representative of Washington, DC 20555 the Town of Rye 155 Washington Road Robert A. Backus Esquire Rye, NH 03870 116 Lowell Street P.O. Box 516 Calvin A. Canney Mancehster, NH 03105 City Manager City Hall Philip Ahrens, Esquire 126 Daniel Street Assistant Attorney General Portsmouth, NH 03801 Department of the Attorney General Augusta, ME 04333 Dana Bisbee, Esquire Assistant Attorney General Mr. John B. Tanzer Office of the Attorney General Designated Representative of 208 State House Annex the Town of Hampton Concord, NH 03301 5 Morningside Drive Hampton, NH 03842 Anne Verge, Chairperson Board of Selectmen Roberta C. Fevear Town Hall Designated Representative of South Hampton, NH 03S42 the Town of Hampton Falls Drinkwater Road Patrick J. McKeon Hampton Falls, NH 03844 Selectmen's Office 10 Central Road Mrs. Sandra Cavutis Rye, NH 03570 Designated Representative of the Town of Kensington Carole F. Kagan, Esq.
RFD 1 Atomic Safety and Licensing Board Panel East Kingston, NH 03827 U.S. Nuclear Regulatory Commission
' * ~
Jo Ann Shotwell, Esquire Assistant Attorney General Mr. Angie Machiros Environmental l'rotection Bureau Chairman of the Board of Selectmen Department of the Attorney General Town of Newbury One Ashburton Place, 19th Floor Newbury, MA 01950 Boston, MA 02108 Town Manager's Office Senator Gordon J. Humphrey Town Hall - Friend Street U.S. Senate Amesbury, Ma.
01913 Washington, DC 20510 (Attn: Tom Burack)
Senator Gordon J. Humphrey 1 Pillsbury Street Diana P. Randall Concord, NH 03301 70 Collins Street (Attn: Herb Boynton)
SEabrook, NH 03874 Richard E. Sullivan, Mayor Donald E. Chick City Hall Town Manager Newburyport, MA 01950 Town of Exeter 10 Front Street
.Exeter, NH 03833 i
' 15.4.3 ROD CLUSTER CONTROL ASSEMBLY MISOPERATION (System Malfunction or Operator Error) 15.4.3.1 Identification of Causes and Accident DescHption Rod cluster co trol assembly (RCCA) misoperation accidents include:
1.
One or more dropped RCCAs within the same group; 2.
A dropped RCCA bank; 3.
Statically misaligned RCCA, 1
4.
Withdrawal of a single RCCA.
Each RCCA has a position indicator channel which displays the position of the assembly. The displays of assembly positions are grouped for the operator's convenience. Fully inserted assemblies are further indicated by a rod at bottom signal, which actuates a local alarm and a control room annunciator.
Group demand position is also indicated.
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i Full length RCCAs art always moved in preselected banks, and the banks are always moved in the same preselected semence. Each bank of RCCAs is divided into two groups of four mechanisms each. The rods comprising l
a group operate in parallel through multiplexing tlyristors. The two groups in a bank move sepentially such that the first group is always within one step of the second group in the bank. A definite schedule of actuation-(or deactuation of the stationary gripper, movable gripper, and lift coils of a mechanism) is required to withdraw the RCCA attached to the mechanism. Since the stationary gripper, movable gripper, _and' lift coils associated with the four RCCAs of a red group are driven in parallel, any single failure which would cause rod withdrawal would affect a minimum of one group. Mechanical failures are in the direction of insertion, or inanobility.
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-51990:1D/102783
The dropped RCCA assemblies, dropped RCCA assembly bank, and statically misaligned RCCA assembly events are classified as ANS Condition II inci-dents (incidents of moderate frequency) as defined in Section 15.0.1.
The single RCCA withdrawal incident is classified as an ANS Condition III event, as discussed below.
No single electrical or mechanical failure in the red control system could cause the accidental withdrawal of a single RCCA fmm the inserted bank at full power operation. The operator could withdraw a single RCCA in the control bank since this feature is necessary in order to retrieve an assembly should one be accidently dropped. The event analyzed must result from multiple wiring failures (probability for single random
- year-refer to Section 7.7.2.2) or failure is on the order of 10
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multiple significant operator errors and subsequent and repeated opera-tor disregard of event indication. The probability of such a combina-tion of conditions is considered low such that the limiting consequences may include slight fuel damage.
Thus, consistent with the philosopPy and format of ANSI N18.2, the event is classified as a Condition III eveEt. By definition " Condition III occurrences include incidents, av one of which'may occur during the i
lifetime of a particular plant", arid "shall not cause more than a small fraction of fuel elements in the reactor. to be damaged..."
This selection of criterion is in accordance with GDC 25 which states,
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"The protection system shall be designed to assure that specified acceptable fuel design limits art not exceeded for av single malfunc-tion of the reactivity control systems, such as accidental withdrawal (not ejection or dropout) of control rods."
(Emphases have been added).
It has been shown that single failures resulting in RCCA bank withdrawals do not violate specified fuel design limits. Moreover, no single malfunction can result in the withdrawal of a single RCCA. Thus, it is concluded that criterion established for the single rod withdrawal at power is appropriate and in accordance with GDC 25.
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5199Q:1D/102783
A dropp;d RCCA or RCCA bank is detectsd by:
1.
Sudden drop in the, core power level as 'seen by the Nuclear Instrumentation System, 2.
Asymetric power distribution as seen on out-of-core neutron detectors or core exit thennocouples, 3.
Rod at bottom signal, 9
4.
Rod deviation alam, or 5.
Rod position indication.
Misaligned RCCAs are detected by:
1.
Asymetric power distribution as seen on out-of-core neutron detectors or core exit themocouples, 2.
Rod deviatica alam, or 3.
Rod position indicators.
The resolution of the red position indicator channel is 15 percent of span (17.2 inches). Deviation of any RCCA from its group by twice this I
distance (10 percent of span, or 14.4 inches) will not cause power dis-tributions worse than the design limits. The deviation alam alerts the operator'to rod deviation with respect to the group position in excess of 5 percent of span.
If the rod deviation alam is not operable, the operator is required to take action as rewired by the technical speci-fications.
If one or som rod position indicator channels should be out of service, l
detailed operating instructions shall be followed to assure. the align-f ment of the non-indicated RCCAs. The cperator is also regired to take action as required by the technical specifications.
l 5199Q:1D/102783 r
Zn the extremely unlikely event of simultaneous elec2rical failures which could result in single RCCA withdrawal, md deviation and rod control urgent failure would both be displayed on the plant annunciator, and the rod position indicators would indicate the relative positions of the assemblies in the bank. The urgent failure alam also inhibits automatic rod motion in the gmup in which it occurs. Withdrawal of a single RCCA by operator action, whether deliberate or by a combination of errors, would result in activation of the same alam and the same visual indications. Withdrawal of a single RCCA results in both posi-tive reactivity insertion tending to increase core power, and an ircrease in local power density in the core area associated with the RCCA. Automatic protection for this event is provided by the overtem-perature AT reactor trip, although due to the increase in local power density it is not possible in all cases to provide assurance that the core safety limits will not be violated.
15'. O. L Plant _ systems and(equipment which are available to mitigate the effects of the various contro1} rod misoperations are discussed in Section 15.0.8 and listed in Tablef13:$s4:3. No single active failure in ary of these
)
systems or equipment wi11 ' adversely iffect the consewences of the acci-dent.
l 15.4.3.2 Analysis of Effects and Consemences 1.
Dropped RCCAs, dropped RCCA bank, and statically misaligned RCCA.
l Method of Analysis a.
One or more dropped RCCAs from the same group.
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For evaluation of the dmpped RCCA event, the transient l.
system response is calculated using the LOFTRAN code. The code simulates the neutmn kinetics, Reactor Coolant System, pressurizer, pressurizer relief and safety valves, pressur-I izer spray, steam generator, and steam generator safety val ves. The code computes pertinent plant variables includ-ing temperatures, pressures, and power level.
51990:1D/102783
Statep31nts art calculated and nuclear mod 21s am used to obtain a hot channel factor consistent with the primary system conditions and reactor power. By incorporating the primary conditions fmm the transient and the hot channel factor from the nuclear analysis, the DNB design basis is shown to be met using the THINC code. The transient J
response, nuclear peaking factor analysis, and DNB design r
basis confimation are perfomed in acconfance with the methodology described in Reference 10.
b.
Statically Misaligned RCCA Stea# state power distribution are analyzed using the com-puter codes as described in Table 4.1-2.
The peaking factors am then used as input to the THINC code to calcu-late the DN8R.
Results One or more Dropped RCCAs a.
Single or multiple dropped RCCAs within the same gmup result in a negative reactivity insertion which may be detected by the power range negative neutmn flux rate trip If detected, the reactor is tripped within circuitry.
approximately 2.5 seconds following the drop of the RCCAs.
The com is not adversely affected duMng this period, since power is decreasing rapidly.. Following reactor trip, nomai
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shutdown pmcedums am followed. The operator may manually retrieve the RCCA by following appmved operating procedures.
For those dropped RCCAs which do not result in a mactor trip, power may be reestablished either by reactivity feed-back or contml bank withdrawal. Following a dropped rod event in manual rod control, the plant will establish a ne.w egailibritan condition. The equilibrium process without l
5199Q:1D/102783
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control system interaction is monotonic, thus removing power overshoot as a concern, and establishing the automatic rod j
control mode of operation as the limiting case.
For a dropped RCCA event in the automatic rod control mode, the Rod Control System detects the drop in power and ini-tiates control bank withdrawal. Power overshoot may occur due to this action by the automatic md controller after which the control system will insert the control bank to restore nominal power. Figures 15.4-13 and 15.4-14 show a typical transient response to a dmpped RCCA (or RCCAs) in automatic contml. Uncertainties in the initial condition are included in the DNB evaluation as described in Reference 10.
In all cases, the minimum DNBR remains above the limit value.
b.
Dropped RCCA Bank A dropped RCCA bank typically results in a reactivity inser-tion greater 'than 500 pc5 which will be detected by the power range negative neutron flux rate trip circuitry. The reactor is tripped within approximately 2.5 seconds follow-ing the drtp of a RCCA Bank. The core is not adversely affected during this period, since power is decreasing rapidly. Following reactor trip, normal shutdown procedures are followed to further cool down the plant. Any action required of the operator to maintain the plant ir, a stabil-ized condition will be in a time frame in excess of ten minutes following the incident.
'c. Statically Misaligned RCCA The most severe misalignment situations with respect to DNBR at 'significant power levels arise from cases in which one RCCA is fully inserted, or where bank D is fully inserted with one RCCA fully withdnwn. Multiple independent alarus, including a bank insertion limit alarm, alert the operator l
5199d:10/102783'
well before the postulated conditions are approached. The bank can be inserted to its insertion limit with any one assembly fully withdrawn without the DNBR f alling below the limit value.
The insertion limits in the technical specifications may vary from time to' time depending on a number of limiting criteria. It is preferable, therefore, to analyze the mis-aligned RCCA case at full power for a position of the control bank as deeply inserted as the criteria on minimum DNBR and power peaking factor will allow. The full power insertion limits on control bank D aust then be chosen to be above that position and will usually be dictated by other criteria.
Detailed results will vary from cycle to cycle depending on fuel arrangements.
For this RCCA misalignment, with bank D inserted to its full power insertion limit and one RCCA fully withdrawn, DNBR does not fall below the limit value. This case is analyzed
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assuming the initial reactor power, pressure, and RCS tem-peratures are at their nominal values including uncertain-ties (as given in Table 15.0-3) but with the increased i
radial peaking factor associated with the misaligned RCCA.
DNB calculations have not been performed specifically for RCCAs missing from other banks; however, power shape calcu-l 1ations have been done as required for the RCCA ejection l
analysis.
Inspection of the power shapes shows that the DNB arid peak kw/ft situation is less severe than the bank D case discussed above assuming insertion limits on the other b'anks ecuivalent to a bank D full-in insertion limit.
l For RCCA misalignments with one RCCA fully inserted, the DNBR does not fall below the limit value. This. case is analyzed assuming the initial reactor power, pressure, and l
l 5199Q:1D/102783 9
RCS temperatures are at their nominal values, including uncertainties (as given in Table 15.0-3) but with the increased radial peaking factor associeted with the mis-aligned RCCA.
DNB does not occur for the RCCA misalignment incident and thus the ability of the primary coolant to remove heat from the fuel rod is not reduced. The peak fuel temperature corresponds to a' linear heat generation rate based on the radial peaking factor penalty associated with the misa11gned RCCA and the design axial power distribution. The resulting linear heat generation is well below that which would cause fuel melting.
Following the identification of a RCCA group misalignment condition by the operator, the operator is required to take action as rewired by the plant technical specifications and operating instructions.
,:[
2.
Single RCCA Withdrawal Method of Analysis Power distributions within the core are calculated using the computer codes as described in Table 4.1-2.
The peaking factors are then used by THINC to calculate the DNBR for the event. The case of the worst rod vitixirawn fmm bank D inserted at the insertion limit, wit'h the reactor initially at f ull power, was analyzed.
This incident is assumed to occur at beginning-of-i life since this results in the minimum value of moderator tem-
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sierature coefficient. This assungtion maximizes the power rise and minimizes the tendency of increased moderator temperature to flatten the power distribution.
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51990:1D/102783.
Rasults For the single rod withdrawal event, two cases have been con-sidered as follows:
a.
If the reactor is in tie manual control mode, continuous withdrawal of a single RCCA results in both an increase in com power and coolant temperature, and an increase in the local hot channel factor in the area of the withdrawing RCCA.
In tenns of the overall system response, this case is similar to those presented in Subsection 15.4.2; however, the increased local power peaking in the area of the with-drawn RCCA results in lower minimum DNBR's than for the withdrawn bank cases. Depending on initial bank insertion and location of the withdrawn RCCA, automatic reactor trip my not occur sufficiently fast to prevent the minimum DNBR fmm falling below the limit value. Evaluation of this case at the power and coolant conditions at which the overtem-
, perature delta T trip would be expected to trip the plant stows that an upper limit for the number of rods with a DNBR less than the limit value is 5 percent.
b.
If the reactor i t in the automat +c centrol mode, the rMti-pie failums that result in the withdrawal of a ringle RCCA
. will result in the irmobility c? the other RCCAs in the contmiling bank. The tri. Ment will then proceed in the same manner as Case r. described above.
For such cases as above, a mactor trip will ultimately ensue, although not sufficiently fast in all cases to prevent a minimum DNBR in the core of less than the limit value. Following mactor trip, normal shutdown procedums are followed.
15.4.3.3 Radiologica1 Consecuences O =. e x M vi3 FSAf{ Se.dow.
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51990:1D/102783 9
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15.4.3.4 Conclusions For cases of dropped. RCCAs or dropped banks, for which the reactor is tripped by the power range negative neutmn flux rate trip, there is no reduction in the manJin to core thermal limits, and consequently the DNB design basis is met.
It is shown for all cases which do not result in reactor trip that the DNBR remains greater than the limit value and, therefore, the DNB design is met.
For al'1 cases of any RCCA fully inserted, or bank D inserted to its rod insertion limits with any single RCCA in that bank fully withdrawn (sta-tic misalignment), the DNBR remains greater than the limit value.
For the case of the accidental withdrawal of a single RCCA, with the reactor in the automatic or manual control mode and initially operating at full power with bank D at the insertion limit, an upper bound of the number of fuel rods experiencing DNB is 5 percent of the total fuel rods in the core.
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Reference 10 - Morita, T., et al., " Dropped Rod Methodology for Negative Flux Rat'e Trip Plants," WCAP-10298-A, June 1983.
This reference should be added to Section 15.4.10.
l 51990:1D/102783
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