ML19322A264
| ML19322A264 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 03/01/1979 |
| From: | Herbein J METROPOLITAN EDISON CO. |
| To: | Reid R Office of Nuclear Reactor Regulation |
| References | |
| FOIA-79-98 GQL-0298, GQL-298, NUDOCS 7903080317 | |
| Download: ML19322A264 (13) | |
Text
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Metropolitan Edison Company e
Post Oflice Box 542 Reading Pennsylvania 19640 215 929-3601 Writer's Direct Dial Number March 1,1979 GQL 0298 Director of Nuclear Reactor Regulations U. S. Nuclear Regulatory Commission Washington, DC 20555 I
Attention:
Mr. R. W. Reid, Chief Operating Reactors Branch No. h
Dear Sir:
Three Mlle Island Nuclear Station, Unit 1 (TMI-1)
Operating License No. DPR-50 Docket No. 50-289 Cycle 5 Reload - Additional Information Your letter of February 16, 1979, requested additional information to support the TMI-1 Cycle 5 Reload.
Enclosed, pleace find Met-Ed's responses to your requests.
Draft responses were transmitted informally to Mr. Dominic Dilanni of your staff on February 20, 1979, in response to your informal trans-mittal of Februar/ 9,1979 In that your letter of February 16, 1979 was not received by Met-Ed until February 26, 1979, and consistent with discussions with your Mr. DiIanni, Met-Ed's responses are being submitted on this date.
Also enclosed, please find a revised Page 2-2 of the TMI-1 Technical Specifications. Page 2-2, which was inadvertently omitted from the origi-nal submittal of December 28, 1978 (GQL 2068), has been revised to include the new values for FAH and Fq.
S cerely, t
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J. G. Herbein
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Vice President - Generation JGH:PJS:c1b Enclosures
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Metropohtan Edison Company is a Memtwr of the General Pubbc UtAl;es System i
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m TMI-l CYCLE 5 RELOAD Additional Information i
1.
NRC CONCERN i
The control rod group withdrawal anticipated operational occurrence analysis shown in the FSAR is based (nominal conditions) on assumed values of the moderator and doppler temperature coefficients less adverse than values of j
these coefficients predicted to occur during the forthcoming cycle.
Sen-sitivity studies shown in the FSAR show the ef fect of the more negative doppler coefficient but do not span the range of the anticipated moderator temperature coefficient.
Please explain how the system trip setpoints l
afford-plant protection (DNBR and RCS pressure) at these more adverse con-ditions.
Please consider the full range of bank worths in your analysis.
~
I l
MET-ED RESPONSE Table 7-1 of the TMI-1 Cycle 5 Reload Report indicates that the h9C valuer I
(worst conditions for the rod withdrawal accident) for the FSAR Moderator j
and Doppler reactivity coefficients are less negative than the Cycle 5 j
values.
Although Figures 14-12 and 14-13 of the FSAR indicate that the i
reactivity coefficients create higher pressures, it must be recognized that these figures are based on a senaitivity study at one given rod j
worth and may not be indicative of a pressure versus reactivity rela-tionship at another rod worth.
The important relation is the peak pres-sure versus rod withdrawal rate, shown in Figure 14-10, which spans the i
range of. rod worth and reactivity feedback combinations, encompassing the reactivity versus power and pressure relations of the Cycle 5 Reload.
l Changes in 'the reactivity coef ficients shif t the point where the peak pressure occurs with respect to rod withdrawal rate and/or move the range i
1 of single to all control rod groups (based on the nominal values listed
(
in Tables 14-3, 14-5, and 14-6) along the curve.
The magnitude of the peak pressure will not change, and therefore, the design over power and/or RCS pressure limits will not be exceeded.
Since the Cycle 5 core parameters do not result in more adverse conditions than the worst cases studies in the FSAR, the system trip setpoints do provide plant protection for DNB and RCS pressure concerns.
2.
NRC CONCERN The dropped rod anticipated operational occurrence analysis shown in the FSAR is based on a doppler coefficient less adverse than the value pre-dicted for the forthcoming Cycle 5.
Explain why you consider the FSAR analysis bounding for Cycle 5.
Provide the post rod drop peak enthalpy rise assumed in the FSAR analysis and predicted for Cycle 5.
MET-ED RESPONSE As nnted in the response to NRC CONCERN 1, the severity of the transient is less dependent of doppler and moderator (moderator even less so, due to the rapid insertion of reactivity) reactivity coefficients than on the maximum worth of the dropped rod.
From Table 7-1 of the Cycle 5 Reload Report comparing the dropped rod worth for the FSAR analysis of 0.46% ok/k with the Cycle 5 value of 0.2% Ak/k, the factor of more than 2 times in difference between the two values indicates by inspection that the FSAR case will bound Cycle 5 for small variations in the reactivity coefficients.
The post rod drop peak enthalpy rise is not an FSAR require-ment.
Babcock and Wilcox has not performed this analysis for any of its plants.
l 1
- 3.
NRC CONCERN The ejected ' rod accident analysis presented in the FSAR is based on values of B ef f less adverse than predicted for Cycle 5.
Explain why you con-sider the FSAR analysis bounding for Cycle 5.
Furthermore, confirm that the post ejected value of the peak linear heat rate assumed in the FSAR analysis bounds the Cycle 5 predicted values.
MET-ED RESPONSE Although the ejected rod accident analysis presented in the FSAR is based on larger B ef f values (0.0071 vs. 0.0058), the FSAR analysis is still bounding for Cycle 5.
This is a result of the larger worth (0.65% Ak/k) assumed for the ejected rod in the FSAR analysis, as compared to 0.25% Ak/k for Cycle 5.
This is best illustrated if the reactivity is expressed in terms of dollars = p/S.
For the FSAR case, this is $ =.0065/.0071 = p/6 0.92.
For Cycle 5, reactivity added is $ =.0025/.0058 = 0.43, more than a factor of 2 less than the FSAR.
The post ejected value of the peak linear heat rate assumed in the FASR analysis bounds the Cycle 5 predicted values, since the design peak is used for the FSAR analysis and the Cycle 5 peaks have been shown to be less.
4.
NRC CONCERN Please confirm the applicability of the BAW 1461, " Reactivity Insertion Assumptions used In Safety Analysis Calculations", to the analysis of TMI-1, Cycle 5.
If applicable, show that sufficient margin will exist
.during Cycle 5 to accommodate the 0.09 DNBR reduction during the hypo-thetical four pump coastdown sited in BAW 1461.
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MET-ED RESPONSE l
The topical report, BAW-1461, " Reactivity Insertion Assumptions used In Safety Analysis Calculations", is applicable to TMI-1, as noted in the introduction of the report.
Since the 4 pump coastdown assumption. are based on the worst case BOC conditions, Cycle 5 and all subsequent cycles are bounded by the information contained in the report.
Sufficient margin will exist during Cycle 5 to accommodate the 0.09 DNBR reduction during the hypothetical 4 pump coastdown, because the minimum DNBR from Table 6-1 of the Cycle 5 reload report is given as 1.98.
This would provide Cycle 5 t
with sufficient margin to accommodate the 0.09 (9 point) DNBR reduction, l
as did the bounding case in the topical.
5.
NRC CONCERN Please confirm that the clad collapse calculation for Cycle 5 were per-formed using the CROV computer code and associated standard modeling techniques.
MET-ED RESPONSE The creep collapse analysis was performed based on the CROV code topical BAW-10084, Rev. 1.
(See TMI-1 Cycle 5 Reload Report, Reference 3).
6.
NRC CONCERN Provide the analytic bases for the revision of Technical Specification 3.2.2 which would increase the minimum boric acid mix tank level from 800 ft.
to 906 ft.3,
MET-ED RESPONSE I
Boric acid storage volumes required for RCS boration to cold shutdown are sensitive to fuel cycle physics parameters such as doppler deficit, mod-i erator deficit, total and stuck rod worths, fuel enrichment, and batch size.
These parameters change on a cycle-to-cycle basis, thereby af fecting the boron concentration requirements for shutdown.
The shutdown calcul-ations require a 1% Ak/k shutdown margin with no xenon and highest worth stuck rod.
The increased boric acid volume for TMI-1 Cycle 5 relative to Cycle 4 is a direct result of increased boron concentration requirements imposed by fuel cycle differences between the two cycles.
7.
NRC CONCERN Cycle 5 values of permitted axial power shape rod (APSR) position vs.
core power, shown as proposed Figure 3.5-2H of the plant Technical Speci-fications, will require long term insertion of the APSR during rated power production.
The APSR is to be withdrawn no less than 6.1%, nor no more than 45%, during operation of greater than 92% of rated power.
Please provide predicted values of FAH and Fq following long term operation with the APSR's 6.1% withdrawn and subsequent withdrawal of the APSR's to 45% withdrawn.
MET-ED RESPONSE The APSR position at full power at which the core of fset will be minimized is approximately 30% withdrawn.
This configuration will be maintained for long term steady state operation.
The limiting APSR positions would only be approached for the control of short-term, transient axial ef fects.
The following table gives the values of Fq (peak pellet) and FAH (peak pin) predicted for the nominal and limiting APSR positions for the end of Cycle 5 after long term operation with the APSR's at 32% withdrawn.
The values below are nominal, no uncertainties have been added.
APSR Position
%WD F6H (Location *)
Fq (Location *)
6.1 1.27 (K-11) 2.17 (L-12) 32 1.28 (K-11) 1.54 (L-14) 45 1.28 (K-11) 1.93 (L-12)
- Locations are 1/8 core symmetric It should be noted that withdrawl of the APSR's to 45% WD without movement of Bank 7 from its nominal position (287% WD Rod Index) will produce an imbalance of -18.8%, which is outside of and would be precluded by the im-balance limits of proposed Technical Specification Figure.
8.
NRC CONCERN Figure 5-1 of your Cycle 5 Reload Report shows the beginning of cycle predicted planar power distribution with the APSR's inserted.
Does this calculation (a two dimensional PDQ07 calculation) represent the APSR's as if they were full length, full strength rods, or have cross sections been adjusted to represent the reduced length of the APSR's?
MET-ED RESPONSE The two dimensional PDQ07 calculations represent the reduced length of the APSR's by having flux and volume weighted cross sections for the APSR's pins from a three dimensional PDQ07 calculations which had the :
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APSR's shown explicitly.
Further normalization of the two dimensional model to the three dimensional analy
- is done by applying an increased
. axial leakage (buckling) to the fuel assemblies containing the APSR's.
This model correctly accounts for the radial peaking and assembly average burn up for the assemblies containing APSR's in the two dimensional cal-culations.
4 9.
NRC CONCERN 2.
Please confirm the applicability of BAW-10121P, "RPS Limits and Setpoints",
I to TMI-1, Cycle 5.
MET-ED RESPONSE Topical Report BAW-10121P, "RPS Limits and Setpoints", was written speci-fically to address RPS - II type plants.
TMI-1 is an RPS-1 plant, and therefore, the information contained in the report does not apply to TMI-1.
The techniques used to determine the RPS setpoints for TMI-1 are outlined I
in Section 2.3, Bases, of the TMI-1 Technical Specifications.
10.
NRC CONCERN f
Please provide the quantitative, rather than qualitative, bases for your revision of the bypass flow to 10.4% of total flow to accommodate the ef fect of orifice rod assembly removal.
MET-ED RESPONSE A detailed review of the methods used to calculate guide tube leakage was conducted by the NRC staff in conjunction with the review of ' Davis -
Besse Cycle -1 operation with BPRA's and ORA's removed This review
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resulted in a may 1978 meeting in Bethesda between NRC's Mr. M. W. Hodges and B&W's Mr. G. A. Moyer. The NRC approved the guide tube leakage cal-culation at that time.
Since then, B&W has used the same method to license Oconee I
, Oconee II
, Oconee III('}, Crystal River 3
, and Rancho Ceco 11.
NRC CONCERN Please provide the quantitative, rather than qualitative, bases of your review of the peak enthalpy rise, FAH, from 1.78 to 1.71 to accommodate the revised bypass flow.
MET-ED RESPONSE The removal of orifice rods in Cycle 5 will increase core bypass flow by 2%.
To offset the reduction in core flow, credit was taken for some of the large margin between the calculated Cycle 5 radial x local peak of 1.403 and the previously used reference design peak of 1.78 (FAH).
A value of 1.71 was chosen for FAH.
The primary impetus for the use of this value was that it had previously been reviewed and approved by NRC for use in the thermal hydraulic design of Davis-Eesse, Cycle 1
, before the concern with BPRA and ORA latching mechanisms.
Therefore, when it became desirable to remove ORA's and/or BPRA's, precedent had already been set for the 1.71 value.
Other B&W plants which have had FAH reduced from 1.78 to 1.71 are Oconee I Cycle 5(2), Oconee II Cycle 4(3), Oconee III Cycle 4 Rancho Seco 3(6), and Crystal River 3 Cycle 1 Experience has shown that 1.71 is more realistic than 1.78 but still provides con-servative margins to steady state and maneuvering peaking limits.
Transient analysis is begun from initial conditions at 102% power.
The minimum DNBR at 102% power has increased from 2.24 to 2.33 in going from w
the 1.78 to 1.71 radial x local peak. The transient analysis applicable to cycle 4 is conservative for cycle 5 because of the increase in initial minimum DNBR.
' The limiting flow transient for TMI-1 is the one pump coastdown which determines the flux / flow trip setpoint.
The cycle 5 flux / flow setpoint is 1.08 for TMI-1.
The minimum DNBR during the one pump coastdown with this setpoint is 1.74 based on the 1.71 value of FAH.
This leaves 20%
i margin to the minimum DNBR criteria for cycle 5 which is 1.43 with 11.2%
1 rod bow penalty.
2 12.
NRC CONCERN Are Figues 8-1 and 8-2, Core Protection Safety Limits, Trip Setpoint for Nuclear Overpower Based on RCS Flow and Axial Power Imbalance, respectively, based on an assumed FAH of 1.71 or 1.787 MET-ED RESPONSE Figues 8-1 and 8-2 are based on an assumed F6H of 1.78.
13.
NRC CONCERN Table 1 of your submittal shows that safety limits calculated for Cycle 5 are less restrictive than the proposed Technical Specifications Safety Limits (SL).
By inference you assert that the Limiting Safety System l
Setpoints (LSSS) corresponding to Technical Specification SL are more restrictive than the LSSS that would correspond to the cycle 5 SL.
Please confirm this assertion.
Consider transient DNBR degradation during the course of postulated transients for which the LSSS are to provide protec-tion, as well as steady state conditions used to determine the SL. _
m
MET-ED RESPONSE The proposed Technical Specifications Safety Limits (Table 1 of the sub-mittal) are based on the Limiting Safety System Setpoints fror. the latter part of Cycle 4 (Figure 8-2 of the Cycle 5 Reload Report).
These limits were determined for the Standard Tech. Specs. by providing the most restrictive envelope, such that all future cycles would be bounded.
Un-3 like previous cycles, the Standard Tech. Spec. Safety Limit envelope was directly calculated from the limits of the cycle 4 trip setpoint envelope, j
thereby providing a more restrictive Safety Limit envelope but allowing I
greater variations in the of fset Limits for subsequent cycles.
The re fo re,.
l as noted in the question, the LSSS corresponding to the Tech. Spec. SL i
are more restrictive than the LSSS corresponding to the Cycle 5 SL noted in Table 1 of the referenced submittal.
With respect to DNBR degradation during the course of postulated transients, the flux / flow trip setpoints provide protection to maintain adequate margin for DNB.
The " winged" portions of each pump operation envelope provide adequate margin to MDNBR for steady state conditions.
i 14.
NRC CONCERN
)
Please commit to provide a startup test report.
MET-ED RESPONSE Within 90 days following completion of TMI-l's Cycle 5 startup and physics testing, a startup and test report will be submitted to NRC. s
y REFERENCES 1.
Attachment i to Application to Amend Operating License for Removal of Burnable Poison Rod and Orifice Rod Assemblies, BAW-1489, Rev. 1, May 1978.
2.
Oconee I Cycle 5 - Reload Report - BAW-1493, Rev. 2, September 1978.
i 3.
Oconee II Cycle 4 - Reload Report - BAW-1491, August 1978.
f l
4.
Oconee III Cycle 4 - Reload Report - BAW-1486, Rev. 1, June 1978.
5.
Crystal River Unit 3 - Licensing Considerations for Continued Cycle 1 i..
Operation Without Burnable Poison Rod Assemblies, BAW-1490, Rev. 1, July 1978.
6.
Rancho Seco Nuclear Generating Station, Unit I - Cycle 3 Reload Report -
BAW-1499, September 1978.
4 7.
Davis - Besse Nuclear Power Station, FSAR.
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O The elevated location where the preccure is actually measured.
The curve presented in Figure 2.1-1 represents the conditions at which a minimum DNBR of 1.3 in predicted for the maximum pocaible thermal power (112 percent) when the reactor coolant flow is 139.8 x 10+6 lba/h, which ic less than the actual flow rate for four operating reactor coolant pumps. Thin curve is based on the following nuclear power peaking factors (2) with potential fuel densification and fuel rod bowing effects; F"
= 2*57' F "!!
N q
A
= 1.71; F
= 1.50 z
The 1.5 axini penking factor annoeinted with the cocine flux chape providen a lencer margin to a DNPR of 1.3 than the 1.7 axial penking factor nncocinted with a lower core flux dictribution. For thic ren"on the cocine flux chape andthenacociatedF3=1.50inmorelimitingandthucthemoreconservative accumption.
The 1.50 cosine axial flux shape in conjunction with FAH = 1.it define the l
reference denign penking condition in the core for operation at the maximum overpower. Once the reference penking condition and the nacociated thermal-hydraulic nituation han been e,tablished for the hot channel, then all other combinations of axini flux nhnpes and their accompanying radials munt recult in a condition which will not violate the previously established design criteria on DNBR. The flux chapen examined include a vide range of positive and negative offcet for steady state and trancient conditionc.
Thene design limit power penking factora are the most rectrictive enlculated at full power for the range from all control rods fully withdrawn to maximum allowable control rod incertion, and form the core DNBR design basis.
The curven of FI ure 2.1-2 are based on the more rectrictive of two thermal C
limita and include the effects of potential h2el densifiention and fuel rod bowing; n.
The 1.3 DNBR limit produced by a nuclear power penking factor of F f=P.57 of the combination of the radial penk, axial p"nk, ilnd ju u l t. ion o f f.h" n x i al penk t hot. yieldn no lecc than 1.3 DNBR.
b.
The combination of radial and axini peak that preventa central fuel melting at the hot spot. The limit is 19.6 kW/ft.
Power peaking in not n directly observable quantity and therefore limitc have been established on the basis of the reactor power imbalance produced by the power peaking.
The specified flow rates for curven 1, 2, and 3 of Figure 2.1-2 correspond to the expected minimum f3ov raten with four pumpc, three pumps, and one pump in each loop, recpectively.
2-2