ML20216J508
| ML20216J508 | |
| Person / Time | |
|---|---|
| Site: | Salem  |
| Issue date: | 09/10/1997 |
| From: | NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML20216J498 | List: |
| References | |
| NUDOCS 9709170342 | |
| Download: ML20216J508 (9) | |
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anyk UNITE 3 STATES 3-B NUCLEAR REGULATORY COMMISSION f
WAaHINGTON, D.C. Siw&4001
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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION ERATED TO AMENDMENT NO. 183 TO FACILITY OPERATING LICENSE NO. DPR-75 PUBLIC SERVICE ELECTRIC & GAS COMPANY PHILADELPHIA ELECTRIC COMPANY DELMARVA POWER AND LIGHT COMPANY ATLANTIC CITY ELECTRIC COMPANY
}ALEM NUCLEAR GENERATING STATION. UNIT NO. 2 DOCKET NO. 50-311
1.0 INTRODUCTION
By letter damt August 19, 1997, as supplemented by letter dated August 20, 1997, the Public Service Electric & Gas Company (the licensee) submitted a request for changes to the Salem Huclear Generating Station, Unit No. 2, Technical Specifications (TSs).
The requested changes would increase the allowable band for control and shutdown rod demanjed position versus indicated position from i 12 steps to i 18 steps when the pewer level is not greater than 85% rated thermal power.
2.0 EVALUATION The analog rod position indication system (ARPI) system is designed to an accuracy of 12 steps. Therefore, in order to guarantee a rod misalignment of less than 24 steps (12 steps misalignment plus 12 steps ARPI uncertainty), the individual ARPI readings must be no larger than 12 steps.
In order to justify changing the misalignment limit to i 18 stus, the licensee did evaluations for misalignments of up to 30 steps (18 steps indicated plus 12 steps uncertainty).
The TS limits on peaking factors F, and F6H increase as the power level lowers. The increase in the limit for F and FAH was used to accommodate the larger than 112 steps misalignment at the reduced power levels. To justify the increase in allowable rod misalignment at a reduced power level, the following were evaluated:
1.
reactivity control 2.
control rod misoperation (dropped rods and static rod misalignments) 3.
rod ejection 4.
power operation with misaligned rods.
The principal tool used in the analysis was the Westinghouse PHOENIX-P/ANC core design system documented in References 2 and 3.
For this analysis the changes in peaking factors rather than the absolute values of the peaking yy () OY 9709170342 970910 DR ADOCK 05000311 PDR
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. factors were of interest.
For each case calculations were performed for misalignments of 124 and 130 steps and compared to the corresponding non-misaligned reference case. The FAH and F compared as a function of axial offset (Ab)for these cases were calculated and throughout the anticipated allowable range of operation. All calculations supporting this report used a HFP A0 band of 115%.
The analysis was performed with two different models of the Salem core, the Unit 2 Cycle 10 core model and a " bounding" future cycle model.
Applicability for each future cycle will be determined during the reload design process.
-2.1 Reactivity Control To demonstrate that reactivity control was acceptable with the additional allowed misalignment, the reactivity effect of a misaligned bank by an additional 6 steps was calculated for both core models at Hot Zero Power (HZP), Hot Full Power (HFP) and part-power conditions. The change was found to be less than 100 pcm.
These calculations were performed for Ead of Cycle (EOC) conditions since that represents the point in cycle with the least available shutdown margin.
For future cycles, if a cycle-specific calculation is not performed, the rod insertion allowance calculated as part of the reload safety evaluation will be conservatively increased by 120 pcm.
2.2 RCCA Misoperation Events The RCCA misoperation events (dropped RCCAs and statically misaligned RCCAs) are events initiated by the movement or displacement of one RCCA rod or bank from its normal position. These events result in reactivity and power distribution anomalies.
A change in the number of steps of n;isalignment allowed does not effect the results of these events since these events bound the misalignment cases.
2.3 Rod Ejection The rod ejection analysis is performed at HZP and HFP, Beginning of Cycle (BOC) and EOC conditions. The physics parameters of interest are the available trip worth following a rod ejection, the ejected rod worth and the post-ejection F,.
Calculations were performed for both core models. The results of these calculations showed that the maximum increases in F, and ejected rod worth were well within the margin on these parameters.
For future cycles if a cycle-specific analysis is not performed the calculated ejected
-rod peak F,t any time in the cycle.will be multiplied by 1.085 to bound the additional 6 s misalignmen Likewise the ejected rod worth will be multiplied by 1.065.
In addition the available trip worth following an ejected rod will be reduced by 100pcm, which bounds the calculated values.
2.4 Power Operation with Misaligned Rod Power distributions with control rod misalignment of 30 steps (18 steps misalignment plus 12 steps for ARPI uncertainty) were evaluated.
To determine
'. the misalignment cases to be analyzed for this technical specification change, an evaluation of the rod control system was performed, drawing from the failure Mode and Effects Analysis.. These analyses were performed to evaluate the impact of RCCA misalignment'on steady state power distribution.
Calculations were performed for both inward and outward misalignments from the demand counter position. Multi)1e misalignments as well as single misalignments were analyzed.
T1e cases analyzed included 800, MOC and EOC cases for both core models. A total of over 200 cases were examined for axial offsets from -15% to +15%.
Comparisons were made between the peaking factors assuming the 18 step misalignment, the 12 step misalignment and the base case (control bank D at rod insertion limit (RIL)).
The results indicate that the maximum incremental increase in F and FM due t1 an additional misalignment of six steps is 3.6%
and 2.4% resp,ectively.
Since the technical s)ecification limits on F and FM for 85% power are 18% and 4.5% greater than tiose at 100% power, the,small changes in F, and FM due to the larger misalignments are adequately accommodated.
2.5 Summary The proposed TS changes modify TS 3.1.3.1, 4.1.3.1, 3.1.3.2, and 4.1.3.2 and associated bases. The changes renlace the rod misalignment value of il2 steps with 118 steps if RTP is not above 85%. The bases have been modified to reflect'the new allowed rod misalignment.
RCCA misalignments up to 30 steps (18 steps indicated plus 12 steps ARPI uncertainty) have been evaluated for impact on peaking factors and reactivity worth. The results with respect to reactivity control, RCCA misoperation events and rod ejection events have been shown to be acceptable.
For power operation with misalignment of 118 steps the results of the analysis showed t1at the incremental increases in the peaking factors were only a small fraction of the increase in the peaking factor limits for power levels less than 85%. Thus it has been shown that the increase in peaking factors will be accommodated at or below 85% of RTP and the change to the technical specification to allow misalignment of up to 18 steps is acceptable.
3.0 STATEMENT OF EXIGENT CIRCUMSTANCES In the August 19, 1997 submittal, the licensee requested that the amendment be reviewed on an exigent basis to provide additional operational flexibility, to allow the orderly resumption of startup and preclude unwarranted power transients. As a result of the rod position indication being at minus 13 steps for demanded position for two rods, Salem Unit 2 completed a TS required shutdown un August 19, 1997.
In the August 20, 1997,. submittal, the licensee stated that, in early August 1997, the licensee, in conjunction with vendor recommendations and participation, revisod the calibration procedures to more closely reflect the
-original Westinghouse calibration procedures. The rod position indication
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- system was.successfully calibrated and Salem Unit 2 went critical on August 17, 1997.= On August 18, during performance of reactor physics testing-(rod swap). two control rods deviated from their group demand counter by 13:
- steps, one step over the limit. As a result, Sales Unit 2 entered TS Limiting Condition.for Operation 3.3.2.1 and shutdown on August 20, 1997.
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Investigation into this' apparent alsalignment did not indicate any deficiencies with the calibration or circuitry. Therefore, prior to August 19, 1997, the licensee could not have foreseen the need to expedite this change. Salem Unit 2 is ex>ected to restart and a similar problem could arise that would necessitate a slutdown.
I Based on the above, the Comission finds that exigent circumstances exist and that the provisions of 10 CFR 50.91(a)(6) apply. The licensee and the Commission must act quickly and time does not permit publication of a Federal Reaiso.er notice allowing 30 days for prior public comment.
Instead, as detal' ed-below, notice was published in local media in the area surrounding the plant. As discussed in Section 4.0, the Comission has determined that the amendment involves-no significant hazards considerations. The Comission also finds, pursuant to 10 CJR 50.91(a)(6)(vi), that the licensee did not l-create the exigency to tvoid the normal notice and coment process.
Accordingly, the Commission published a public notice of the proposed amendment, issued a proposed finding of no significant hazards consideration and requested that any comments on the )roposed no significant hazards-i.
consideration be provided to the staff )y the close of business on.
L September 3,1997, pursuant to 10 CFR 50.91(a)(6). This notice was published
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in the Wilmington News Journal on August 22, 1997, and will be in Today's Sunbeam on August 24, 1997.
4.0 ColMENTS During the comment period, the Comission received telephone calls from two individuals and a letter and telephone call from a third individual. The following is a summary of the comments that were received.
One individual asked several questions: (1) How many steps are there from j
fully out to fully in?; (2) Have any other plants received a similar-amendment?; (3) What is the basis for the 12 step difference that is currently
. allowed in the Technical Specifications; and (4) What is the rush to process the amendment?
The staff provided the individual with the following responses: (1) Full out-to full in is 228 steps.
(2) Similar amendments have been granted for Turkey Point Units 3 and 4,:and North Anna, Units 1 and 2. -(3D The 12 steps is the allowed misalignment at 100% power. At lower. power levels, there is more amargin available and therefore a larger misalignment is permitted. The amendment allows-a misalignment of 10 steps at power levels not greater than 85%. (Additional discussion is provided in Section 2.0)
(4) The amendment was processed on an exigent basis to preclude an unnecessary plant shutdown.
Section 3.0 provides additional discussion on the need for the exigent j
amendment.
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I i A second individual commented that the amendment appeared to be as though the NRC was " slackening the rules.' The staff responded that the amendment was carefully reviewed by the staff's technical experts and it was found that granting the amendment would not have an adverse impact to the health and safety of the public.
The third individual telephoned and sent a letter with his coments.
He concurred with the staff's assessment that the amendment will not adversely affect safety margins at Salem, but disagreed with the need to process the amendment on an exigent basis. On September 3, 1997, the staff spoke to the individual regarding his concerns.
As explained in Section 3.0, the licensee expected that the rod position
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misalignments would be within the Technical Specification limit after achieving critically since they had been within the limit in the last calibration performed on August 17. Accordingly, the licensee could not have foreseen the need to expedite the amendment prior to August 18, 1997, when it was discovered that two rods were outside the Technical Specification limit.
The third individual also referred to LER 50-327/96011, submitted for Sequoyah Nuclear Plant, Unit 1.
This LER describes a situation where the rod position indication system was more than 12 ste:s different than the demand step counter for two control rods.
Segaoyal dealt with the situation by dilution of the reactor coolant system and insertion of the two rods to the point where i
the non-linear response of the rod )osition indication system was less pronounced. By doing this, Sequoyal was able to return within the 12 step difference allowed by the Technical Specifications. The individual asked why Salem could not take the same approach and therefore not need the amendment on an exigent basis.
Sequoyah had been at 15% power and toward the end of the fuel cycle, while Salem was conducting low power physics testing at the beginning a new fuel cycle. Low power physics testing involves determining the worth of each rod in which the position of each rod is important.
Dilution of the reactor coolant system, as done at Sequoyah, would not be permitted during this phase of the testing at Salem. Therefore, it was not appropriate for Salem to reposition the rods as did Sequoyah, to the point where the non-linearity of the rod position system was less pronounced.
5.0 EIML NO SIGNIFICANT HAZARDS CONSIDERATION DETERMINATION The Comission has provided standards for determining whether a significant hazards consideration exists (10 CFR 50.92(c)). A proposed amendment to an operating license for a facility involves no significant consideration if operation of the facility in accordance with the proposed amendment would not (1) involve a significant increase in the probability or consequences of an accident previously evaluated; or (2) create the possibility of a new or different kind of accident from any accident previously evaluated; or (3) involve a significant reduction in a margin of safety.
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. The following evaluation was provided by the licensee:
1.
The proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.
The proposed change to the rod misalignment criteria of (plus or minus) 18 steps for core powers equal to or below 85% of RATED THERMAL POWER (RTP) does not increase the probability of previously evaluated accidents.
Increasing the magnitude of the allowed control rod misalignment is not a contributor to the mechanistic cause of an accident evaluated in any accident analysis.
The magnitude of control rod indicated misalignment is a parameter used to establish the initial conditions for accident evaluation.
The proposed increase in the allowable rod misalignment from the current (plus or minus? 12 steps for reactor powers equal to or less than 85% RTP does not 'nvolve a significant increase in the consequence of any previously evaluated accident.
Rod misalignment affects power distribution, shutdown margin and the ejected rod accident. An extension of the allowable rod misalignment above and below 85% RTP has been analyzed in Westinghouse WCAP-14672. As provided in WCAP-14672, above 85% the allowable misalignment is governed by the available peaking factor margins as determined by flux maps.
PSE&G is simplifying the proposed change by keeping the currently allowed -(plus or minus) 12 step misalignment in Technical Specifications 3.1.3.1 and 3.1.3.2.1 for reactor power greater than 85% RTP.
The PSE&G proposed change is to allow [plus or minus) 18 steps misalignments in Technical Specifications 3.1.3.1 and 3.1.3.2.1 for reactor power less than or equal to 85% RTP. As demonstrated in WCAP-14672, for reactor powers less than 85% RTP, the available peaking factor margin increases faster than any penalty associated with a [plus or minus) 18 step misalignment.
As described in Section 4.0 of the Westinghouse WCAP, a conservative penalty factor has been applied to the rod insertion allowance (RIA) of the shutdown margin calculation to account for rods misaligned an additional [plus or minus) 6 steps (for a total of (plus or minus) 18 steps). Th1s conservative penalty factor is applied as part of the reload analysis in order to satisfy Technical Specification 3.1.1.1.
In addition to the normal, or Condition 1, operational transients, the impacts of increased rod misalignment on Condition II, III and IV accident analysis have also been evaluated. The proposed increase in-rod misalignment does not have a significant effect on any moderator or Doppler reactivity coefficients or defects, boron worth or reactor kinetics parameters.
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To account' for the potential increase in ejected rod parameters,.
conservative penalty factors have been applied to the reload safety
-evaluation >to cover the additiona1 {plus or minus) 6 step _
misalignment. Margin is available in the reload safety analysis to accommodate.this impact.
i Therefore, the proposed amendment does not increase the probability or consequences of any accident previously evaluated 2.' The proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.
No new accident scenarios, failure mechanisms or limiting single failures are introduced as a result of the proposed change to the rod misalignment criteria of (plus or minus118 steps below 85% RTP. The 4
-implementation of the proposed rod misalignment criteria will have no adverse effect on the performance of any other safety related system.
Therefore, the proposed amendment does not create the possibility of a new or different kind of accident from any previously evaluated.
3.
The proposed change does not 'ovolve a signifi' cant reduction in a margin of safety, Operation of the facility in accordance with the proposed amendment would not involve a significant reduction in the margin of safety.
The Technical Specifications allowed increase in peaking factors as a:
power is reduced accommodates the peaking factor penalty associated with the additional [plus or minus]-6 step misalignment for core powers-equal to or less than 85% RTP. Therefore, there is no change
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-to the peaking factors assumed in the safety analysis.
In addition to peaking factors, there is no change in any other current limit input into the safety analysis. As the input, or initial conditions, of the safety analysis have not changed, there is no reduction in the margin to safety.
In addition, the staff concludes, with respect to the second standard, that no physical modifications are being implemented in the facility.
The NRC staff has reviewed the licensee's analysis and based on this review, it appears that the three standards of 10 CFR 50.92(c) are satisfied.
Therefore, the Commission finds that the amendment request involves no
- significant hazards consideration.
6.0 STATE CONSULTATION
-In accordance with' the Commission's regulations, the New Jersey State official was notified of the proposed issuance of the amendment.
By telephone call on August = 21,1997,~ the State official asked whether power measurement
. uncertainties had been considered since the amendment only changes the k..
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. allowable band to 118 steps when power level is not greater than 85% rated thermal power. As explained in Section 2.0, there is adequate margin in the analysis at 85% rated thermal power to account for power measurement uncertainties.
7.0 ENVIRONMENTAL CONSIDERATION
The amendment changes a requirement with respect to installation or use of a facility component located within the restricted area as defined in 10 CFR -Part 20 and changes surveillance requirements. The NRC staff has determined that the amendment involves no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no si occupational radiation exposure.gnificant increase in individual or cumulative The Comission has found that the amendment involves no significant hazards consideration.
Accordingly, the amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9).
Pursuant to 10 CFR 51.22(b) no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendment.
8.0 CONCLUSION
The Comission has concluded, based on the considerations discussed above, that:
(1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in comnliance with the Comission's regulations, and (3) the issuance of the amendment will not be inimical to the comon defense and security or to the health and safety of the public.
Principal Contributor:
M. Chatterton Date: September 10, !997
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. '8.0 BEFERENCES Aug st
'fgg7Public Service Electric and Gas Company, to NRC, dated 2.
T.- Q.- Nguyen, et al.,- Qualification of the PHOENIX-P/ANC Nuclear Design System for Pressurized Water Cores, WCAP-Il596-P-A, June 1988.
4 3.
Y. S. Liu, et al., ANC: A Westinghouse Advanced Nodal Computer Code, WCAP-10965-P-A, December 1985.
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