ML19269E357
| ML19269E357 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 05/12/1979 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19269E356 | List: |
| References | |
| NUDOCS 7906270259 | |
| Download: ML19269E357 (68) | |
Text
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UNITED STATES fi
-t NUCLEAR REGULATORY COMMISSION
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WASHINGTON, D. C. 20555 3.'
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S._AFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT NO. 52 TO FACILITY OPERATING LICEllSE N0. DPR-65 NORTHEAST NUCLEAR ENERGY COMPANY, ET AL MILLSTONE NUCLEAR POWER STATION, UNIT NO. 2 DOCKET NO. 50-336 i
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I 2 51 122 2908 22og-
TABLE OF CONTENTS.
Subject Page 1.0 I n t ro d uc t i o n...........................................
1 2.0 Evaluation.............................................
4 2.1 Cycle 3 Core......................................
4 2.1.1 F uel De s i g n................................
5 2.1.2 Nuclear Analyses Methodology...............
6 2.1.3 Nuclear Parameters.........................
6 2.1.4 Thernal Hydraul ic Properties...............
9 2.2 Pea king Factor Uncertainties......................
11 2.3 Burnable Poison Pin Performance...................
11 2.4 CEA Guide Tube Integrity..........................
12 2.5 Plant Systems Capability and Modifications........
14 i
2.5.1 LOCA Credit for Charging Pump Flow.........
14 2.5.2 RCP Speed Sensing RPS Trip.................
18 l
2.6 Analyses of Anticipated Operational Occurrences...
22 i
2.6.1 Safety Analysis Parameter Values...........
24 l
2.6.2 Transient Analysis Input to the TM/LP Setpoint.................................
25 2.6.3 Computation of Required Overpower Margin i
from Transient Analyses..................
26 2.6.4 C EA W i th d ra wa 1.............................
27 2.6.5 RCS Depressurization..........-............
27
~
~2.6.6 B o ro n 0 i l u t i o n.............................
27 2.6.7 Lo s s o f Lo a d...............................
23 2.6.8 Lo s s o f Fee dwa ter Fl ow.....................
28 2.6.9 Excess Load................................
29
- 2. 6.10 Lo s s o f RCS F10w...........................
29 2. 6.1 1 C E A D ro p...................................
30 2.6.12 Transients Resulting from Malfunction o f One Steam Genera tor...................
30 g
r 2131 123 I
Page 2 TABLE OF CONTENTS (Cont'd)
Page Subject j
l 31 l
2.7 Postul ated Accidents Other than L0CA.................
31 2.7.1, Steam Generator Tube Rupture..................
31 2.7.2 Main Steam Line Rupture.......................
32 2.7.3 CEA Ejection..................................
32 2.7.4 Seized Rotor..................................
33 2.8 Eval uation of LOCA Analysi s..........................
34 2.8.1 Rupture Strain Model Review...................
34 2.8.2 Large Break L0CA..............................
34 2.8.3 Small Break L0CA..............................
36 2.9 Meteorological Data Review...........................
37 2.10 Radiological Consequences of Postulated Accidents....
2.10.1 LOCA, Fuel Handling and CEA Ejection 37 Accidents...................................
2.10.2 Main Steam Line Rupture, Steam Generator Tube Rupture and Waste Gas Decay Tank Failure....
37 2.10.3 Fuel Handling Accident Inside Containment.....
37 l
2.10.4 Engineering Safety Features Component 39 Leakage Outside Containment.................
40 2.10.5 Control Room Habitability After LOCA..........
2.10.6 Conclusion on Radiological Consequences 40 of Postulated Accidents.....................
40 2.11 Radioactive Waste Management.........................
41 2.12 Neutron Shie1d.......................................
2.13 Containment Air Recirculation System Response 45 Time...............................................
4c 2.14 Main Steam Line Safety Va1ves........................
47 2.15 Environmental Qualification of Class IE Equipment....
47 2.16 Containment Electrical Penetrations..................
50 2.17 Steam Generator Surveillance.........................
51 2.18 Piping and Support Systems...........................
2131 124
i Page 3 TABLE OF CONTENTS (Cont'd)
Subject Page 3.0 Technical Specifications...............................
53 4.0 Physics Testing...................................,....,
56 5.0 Appraisal of Operating Ef fectiveness...................
56 6.0 Conclusions............................................
57 6.1 Con cl u s io n o n S a fe ty..............................
57 6.2 Power Increase....................................
57 6.3 Environmental Considerations......................
57 TOPICAL REFERENCES..........................................
58 USE OF TOPICAL REP 0RTS......................................
60 LETTER REFERENCES...........................................
62 ENCLOSURE - Memorandum, " Millstone-2 Stretch Power Request, j
G. Klinger to B. Grimes, February 9,1979.......
65 t
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1.0 Introduction 16, 1977, December 15, 1978, February 12, 1979 By applications dated December and March 2,1979 and supplemental information as listed in the reference section, Northeast Nuclear Energy Company (NNECO or the licensee) requested an amendment to Facility Operating License No. DPR-65 for the Millstone The amend-Nuclear Power Station, Unit No. 2 (Millstone-2 or the facility).
l ment request consists of:
Appendix A (Safety) Technical Specifications (TS) changes resulting o
from the analyses of the Cycle 3 reload fuel; Appendix A and Appendix B (Environmental) TS changes required to authorize I
l e
operation of the facility at power levels up to 2700 megawatts thermal I
(MWt);
Continued approval to operate with modified (sleeved and reduced flow) i l
e Control Element Assembly (CEA) guide tubes; Approval of a Reactor Coolant Pump (RCP) Speed Sensing System trip e
in the Reactor Protection System (RPS) to protect the core in the event of the complete loss of RCP flow; f
Approval to take credit for charging pump flow in the Small Break Loss l
e of Coolant Accident (LOCA) analysis; i
I Evaluation of the Neutron Shield installation; e
Addition of a Containment Air Recirculation System response time in the i
e Appendix A TS; i
Approval to operate at reduced power level with inoperable Main Steam s
Line safety valves in accordance with an Appendix A TS; and Information necessary to prepare an Environmental Impact Appraisal (EIA) e evaluating the environmental effects of authorizing operation of the facility at 2700 MWt.
The associated specific TS changes are described in Section 3.0 of the following Safety Evaluation (SE).
In addition this SE addresses our evaluation of:
e New meteoroligcal data; Fuel handling accident inside containment; e
I Engineering safety features component leakage outside e
centainment; i
Control room habitability after postulated loss-of-coolant e
accident; Environmental qualifications of Class lE equipment; e
7 I
e Containment electrical penetratiors; I
Steam generator surveillance; and e
2131'l26, l
e Piping.and support system.
In the latter half of 1978, NNECO indicated their intention to apply for a power level increase from 2560 MWt to the Final Safety Analysis Report (FSAR) ultimate possible power level of 2700 MWt (Ref S. 5,8,11).
The FSAR points out that:
e Physics and core thermal hydraulic analysis is based upon a core power level of 2560 MWt; e Site parameters and other major systems and components, including the engineered safety features and the containment structures, have been evalbated for operation at a core power level of 2700 MWt; and e Certain of the postulated incidents considered in Chapter 14 are evaluated at the higher power level (2700 MWt).
Included in the application for the Cycle 3 reload as an integral part, is the request for a 5.5 percent power level increase (Refs.14,18).
The NRC staff Safety Evaluation Report (SER), dated May 10, 1974, does not specifically mention that our evaluation was performed for a power level of 2700 MWt although the basis for the SER is the licensee's FSAR.
For this reason, system equipment capability including modifications, incident analyses, radiological consequences of accidents, and other related licensing actions will be addressed in this SE in addition to the normal evaluation performed for a cycle reload.
In the NNECO analysis for Cycle 3, performed primarily by Corabustion Engineering (CE), the following changes from the Cycle 2 analysis are identified as:
e Use of ROCS coarse mesh neutronics code to predict core parameters; e Reduction of the assumed measured peaking factor uncertainties; e Adoption of the TORC thermal hydraulic code; Adoption of the CE-1 Departure from Nucleate Boiling Ratio (DNBR) e correction; e Selection of increased radial peaking factors; and e Higher core enrichment to allow increased power level and extended length of cycle.
In portions of the analysis, NNECO and CE have used the Cycle 2 analysis as a " Reference Cycle". Operation for Cycle 2 was evaluated and approved by our Reference 4 SE. The core related i
evaluations are presei.ted in Sections 2.1, 2.2 and 2.3 of this SE.
2131 127 1
1
. In December 1977, during the Cycle 2 refueling of Millstone-2, NNECO identified a severe CEA guide tube wear problem. The temporary repair, approved by Reference 4, was the sleeving of all fuel assemblies to be placed in CEA locations and the sleeving of other worn fuel assem-blies in non-CEA locations to regain safety margins. This same temporary repair has been used at other facilities designed by CE.
In the last refueling of Calvert Cliffs Unit No. 2, we approved the installation of a 16 fuel assembly demonstration test with reduced flow through the CEA guide tubes (Ref. 7).
NNEC0 has proposed to install a four fuel assembly demonstration test of reduced CEA guide tube flow assemblies for Cycle 3 operation.
Section 2.4 of this SE evaluates the CEA guide tube wear problem at Millstone-2.
NNEC0 has proposed a system modification and the new use of an existing system to reach the higher power level.
System capabilities, the new RCP Speed Sensing System, and the use of the charging pumps as an Emergency Core Cooling System (ECCS) are evaluated in Section 2.5 of this SE.
The reanalyses of cll transient and accidents ~ are evaluated in Sections 2.6, 2.7 and 2.8 of this SE.
These reanalyses, in conjunction with the normal operating analyses, are used to compute the setpoints used in the TS.
NHEC0 states that the Cycle 3 proposed setpoints offer the same degree of protection at 2700 MWt as the Cycle 2 setpoints offered at 2560 MWt.
Sections 2.9 and 2.10 of this SE evaluate new meteorological data and the radiological consequences of postulated accidents at 2700 MWt.
The radioactive waste management system is reviewed in Section 2.11 to assure adequacy for operation at 2700 MWt.
NNEC0 has been investigating possible shield designs to reduce the neutron exposure in the containment since the facility startup. A permanent neutron shield will be installed to cover the reactor cavity annulus during the Cycle 3 refueling outage. Our evaluation of this neutron shield is contained in Section 2.12 of this SE.
Sections 2.13 and 2.14 evaluate the proposed changes in the TS for the Containment Air Recirculation response time and the operation at reduced power level when Main Steam Line Safety Va?ves are inoperable. The environmental qualifications of Class lE equipment are addressed in Section 2.15.
Section 2.16 evaluates the containment electrical penetrationifor Cycle 3 operation.
The steam generator surveillance program and results are reviewed in Section 2.17.
Section 2.18 reviews the analyses of some safety related piping systems and supports.
~
2131 128
4-2.0 Evaluation i
In this evaluation of a cycle reload and power level increase for Millstone-2, considerable use is made of generic reviews of various 1
A table at the end of this SE entitled "Use of topical reports.
Most Topical Reports" Summarizes the use of such generic reviews.
In 1
c7 the topical reports have received formal NRC staff approval.
l all cases where a topical report has not received approval, the report has been examined, its methods judged to be reasonable, and an appra.isal i'
has been made that a complete review will not reveal the methodology to On this basis, all topicals referenced are be significantly in error.
judged to be acceptable for this reload and power level increase evalu-l ation.
2.1 Cycle 3 Core During the Cycle 3 refueling outage of the Millstone-2 core, 72 batch B fuel assemblies will be discharged and replaced with new batch E and E* fuel assemblies. The pertinent characteristics of the Cycle 3 core are:
Initial Initial Average Number Shim Total Assembly Number of Enrichment Burnup of Loading Total Fuel Designation Assemblies wt% U-235 MWD /MTU Shims wt% B4C Shims Rods i
i B+
5 2.33 26,000 12 2.7 60 820 0
7,040 I
C 40 2.82 20,200 0
+
C+
16 2.82 25,400 12
.83 192 2,624 r
C 12 2.82 25,400 12
.46 144 1,968 D
48 3.03 8,100 0
0 8,448 i,
D*
24 2.73
.11,200 0
0 4,224 l
0 8,448 E
48 3.24 0
0 0
4,224 E*
24 2.73 0
0 396 37,796 217
+ and
- indicate differences in fuel assembly design of a particular Note:
batcn.
ill31 129
2.1.1 Fuel Design The fresh batch E and E* fuel assemblies are essentially identical to the batch D fuel assemblies used in Cycle 2, with the exception that the batch E fuel assemblies have slightly higher enrichment and slightly lower backfill pressure to compensate for the enrich-ment difference.
The fuel management pattern was developed to accommodate a Cycle 2 The actual endpoint exposure range of 8,100 MWD /MTU to 9,300 MWD /MTU.
core exposure achieved during Cycle 2 was 9,232 MWD /MTU bringing the core After the core average End of Cycle (E0C) exposure to 19,000 MWD /MTU.
reload, the Beginning of Cycle (B0C) :3 core exposure will be 10,500 MWD /MTU making the predicted E0C 3 average core exposure about 20,800 MWD /MTV.
The clad colla ggetime was computed by NNEC0'using the CEPAN code (Ref. p).
This analysis predicted the following clad collapse schedule:
Fuel Cycles Actual or Expected Predicted Time to Batch In Core Operation - EFPH*
Clad Collapse - EFPH*
=
8 1, 2, 3 26,790
>30,000 C
1, 2, 3 26,790
>30,000 0
2,3,4 24.050
>29,000 E
3, 4, 5 25,230
>28,600
- Effective Full Power Hours (24 EFPH = 31.0 MWD /MTU)
We find that these margins are adequate to assure that no clad collapse will occur during Cycle 3 operation.
The FATES code is used to compute all steady state fuel parameters which are required as initial conditions for the LOCA transient codes (Ref. j).
Using the FATES medel, the thermal performance of the various types of fuel assemblies has been evaluated with respect to their Cycle 1 and Cycle 2 burnups, prop'osed burnups during Cycle 3, their respective fuel geometries and expected flux levels during Cycle 3.
NNEC0 has found that the twice burned assemblies have been determined to be the limiting fuel batch with respect to stored energy (hotter fuel pin and lower gap conductance).
The metallurgical requirements of the fuel cladding and the fuel assembly structural members for the batch E fuel have not been changed from the original Cycle 1 and Cycle 2 designs.
Therefore, NNEC0 finds the chemical or metallurgical performance of the batch E fuel will be unchanged from that of the original core fuel and,'th.erefore, discussions in the FSAR, are still valid.
2131 130
. 2.1.2 Nuclear Analyses !1ethodology
~
' The Nuclear Design Model used in previous cycles has been PDQ, PDQ a two-dimensional diffusibn ccde using four energy groups.
l has been accepted industry wide (Ref, t).
For Cycle 3, CE performed the calculations of certain parameters using the ROCS code instead of PDQ.
Using a higher order differencing methodology than PDQ and only one and a half energy groups, ROCS is able to j
compute many parameters nearly as accurately as PDQ in three dimensions I
I with more reasonable computer run time.
For Cycle 3 the following safety parameters were computed using the ROCS code:
- Fuel Temperature Coefficients
- Moderator Temperature Coefficients
- Inverse Boron Worths
- Critical Boron Concentrations
- CEA drop distortion factors and reactivity worths
- Reactivity Scram Worths and A110wences
- Reactivity worth of regulating CEA banks
- Changes in 3-D core power distributions that result from inlet temperatures ma1 distributions (asymmetric steam generator transient)
None of these parameters requires the detailed knowledge of pin powers l
NNEC0 states that in most cases their normally computed by PDQ.
parameters are calculated more accurately by ROCS because of its ability NNEC0 has also stated that to account for three dimensional effects.
they observe guidelines to evaluate the adequacy of ROCS for computing these parameters on a case by case basis.
If ROCS is judged to be not
~
adequate for a certain computation, then the computation is repeated l
using PDQ.
Based on our review we find the use of ROCS to be acceptable for this l
reload.
I l
2.1.3 Nuclear Parameters f
NNEC0 presents their proposed Cycle 3 loading pattern in Reference 18.
Changes in the nuclear parameters are indicated in the following table.
2131 131
=
! I f
I REFERENCE UNITS CYCLE 2 __
CYCLE 3 Dissolved Boron Critical Boron Concentration TCEAs withdrawn) hot, full power, equilibrium PPM 660 830 xenon, BOC Boron Worth 88 93 PPM /% 3p Full Power, BOC 77 82 Full Power, E0C PPM /% 3p I
Reactivity Coefficients (CEAs I
Withdrawn) l Moderator temperature coefficients
-4 hot, full power, equilibrium 10 ap /F
.6
.?
-4 hot, full power, EOC 10 ap /F
-2.0
-1.8 r
f Doppler coefficient
-5 10 ap/F
-1.44
-1.44 f
hot, B0C, zero power
-5 10 ao /F
-1.13
-1.13 hot, B0C, full power
-5 10 Ap /F
-1.22
-1.22 i
hot, E0C, full power Total Delayed Neutron Fraction, 6_ef
.00608
.00624 B0C l
.00526
.00524 f
E0C i
Neutron Generation Time, t
-6 10 sec 28.0 27.2 BOC
-6 10 sec 32.2 31.8 EOC 2131 132
8-Many of the changes in these parameters are quite small.
The increase in the critical boron concentration listed in the table is the result of the higher average enrichment necessary for ir. creased power level operation and an eatended cycle length.
Cycle 3 Moderator Temperature Coefficient (MTC) is calculated to be
-0.2 E-4 ac/F fir Beginning of Cycle (B0C) and -1.8 E-4 ap/F fcr End cf Cycle (E0C).
ihese values are bounded by the values used in the safety analyses for the Reference Cycle. The Doppler Coefficient for Cycle 3 is identical to the value used in the Referen'ce Cycle which is a best estimate value expected to be accurate to within 15 percent.
In order to assure that a conservative value was used in the safety analysis, a value 15 percent greater or less than this was usually used.
We find the values of the MTC and Doppler Coefficients to be acceptable.
At E0C 3, the reactivity worth with all CEAs. inserted assuming the highest j
worth CEA is stuck out of the core is 7.5 percent ap.
The reactivity worth required for shutdown, including the contribution required to control the Steam Line Rupture incident at E0C 3 is 6.1 percent op.
The uncertainty allow-ance is the di~f ference between these two numbers,1.4 percent op. This value corresponds to about 20 percent of the total bank worth which is twice the estimate of the uncertainty in nuclear calculations. The shutdown margin and safeguards allowance is 3.2 percent for both BOC and E0C of Cycle 3 and the reference cycle.
For all situations other than moderator cooldown incidents, the availeble shutdown margin is 3.2 percent ap.
We find these shutdown margins to be acceptable.
The augmentation factor (used to account for the power density spikes due to axial gaps caused by fuel densification) was calculated for Cycle 3 using the nethodology described in Reference j which has been approved by the NRC staff.
These augmentation factors are included in the determination of Fx.
The Cycle 3 maximum augmentation factor of 1.054 is only slightly hi her than the Cycle 2 value of 1.045.
Radial power peaking has increased from Cycle 2.
The safety analysis, which assumes the maximum radial power peaks expected to occur throughout the cycle, was performed with the unrodded Fr of 1.598 and Fxy. of 1.584.
For Cycle 2 the corresponding values were Fr of 1.440 and Fxy of 1.540.
Since these higher peaks for Cycle 3 are reflected in the safety analysis and specified TS limits, we find the use of slightly higher power peaking acceptable.
In the Reference 4 SE, we found that introducing of stainless steel sleeves into the CEA guide tube had minimal affect on reactor physics.
The operation of the Millstone-2 for one cycl,e with all CEA guide tubes sleeved has borne out this conclusion.
0 e
2131 133
In the SE supporting the Cycle 2 reload for Calvert Cliffs Unit No. 2 (Ref. 7), we approved a demonstration test consisting of 16 fuel NNEC0 has proposed a assemblies with reduced CEA guide tube flow.
four-fuel assembly demonstration test of such fuel assemblies.
They f
anticipate no substantial change in axial and radial power distribution as a result of the decreased flow in the modified CEA guide tubes.
f This demonstration test will be discussed in Section 2.4 of this SE.
2.1.4 Thermal Hydraulic Properties In previous cycle analyses, NNEC0 has used the COSMO-INTHERMIC thermal hydraulic codes and the W-3 DNBR correlation (Ref. m).
For the W-3 correlation the 95/95 confidence / probability 1.imit for not suffering Departure from Nucleate Boiling (DNB) is a DNBR greater than 1.30.
In the Cycle,3 analysis NNECO has used the' TORC thermal hydraulic code and the CE-1 DNBR correlation (Refs. m, n), both of which have been approved by the NRC staff.
For the CE-1 correlation the 95/95 confidence /'
probability limit is not violated if DNBR is maintained greater than 1.19.
All normal transient.and accident analyses used to generate the TS limits have been modified to reflect this change.
We find this change in method-ology acceptable.
Since control continues at the 95/95 limit, our accept-ance criteria have not changed, and we find there is no degradation in j
demonstrated DNBR protection from Cycle 2 to Cycle 3.
1 CE has stated that comparisons between TORC-CEl and COSM0-INTHERMIC-W3 demonstrate that COSM0-INTHERMIC-W3 is conservative relative to TORC-CEl j
by from three to approximately eight percent depending on core conditions.
A portion of the DNBR Limiting Conditions for Operation (LCO) and Limiting Safety System Settings (LSSS) operating margin necessary for the power level increase is obtained by using the TORC-CEl analysis.
f l
Recently NNECO has adopted the new CE Thermal Margin / Low Pressure (TM/LP) methodology (Refs. a, r). The change in methodology consists of treating certain relationships between parameters exactly rather than replacing these relationships by conservative linear approximations, as was done in the old methodology.
CE states that the new methodology is more precise I
although less conservative than the old methodology.
P have approved the use of the new TM/LP methodology on other recent relouds (Ref. 7). This i
l new methodology, affecting the TM/LP trip setpoints, allows some additional margin for the power level increase.
i 1
i f
2131 134 r
l i
I j i
t In the DNB limit analysis, the assumed uncertainties in various measured i
parameters are not combined in a single ecuation but are factored into This l
functional relationships as biases at various points in the analysis.
biasing of functional relationships throughout the analysis is equivalent to adding the absolute power unpertainties equivalent to the uncertainties in the various measured parameters and applying the total power uncertainty NNECO has provided the following specific to the toest estimate calculation.
Cycle 3 uncertainties along with their equivalent power uncertainties:
Equivalent Uncertainty Power Uncertainty Axial Shape 0.06 ASIU 1 2.2%
Index (ASI)
Pressure 22 PSI 1 0.8%
Temperature 2F 1 0.9%
Flow 4%
> 5.0%
Power 57 (LSSS)
> 3.5%
2% (LCO)
> 1.4%
In the Cycle 3 analyr,is the equivalent sum of these uncertainties is 12.4 percent for the DNBR LSSS and 10.3 percent for the DNBR LCO.
NNECO believes that these components are statistically independent. We find this acceptable.
This being the case, the proper method for combining these uncertainties is root sum square (RSS).
The RSS combination yields 6.6 percent for the LSSS and 5.8 percent for the LCO, giving a net conser-vatism in the analysis of 5.8 percent for the LSSS and 4.5 percent for the LCO.
For the Cycle,3 analysis NNEC0 has proposed a partial credit of 3 percent for the LC0 and LSSS.
We find this acceptable.
In previous analyses of CE plants this same ;redit has been approved to offset the Fr measurement uncertainty.
The NNEC0 treatment differs from this in that the 3 percent c;redit has been applied instead to adjusting the DNBR ASI LSSS and LC0 limits and to adjust some intermediate variables in the computation of the TM/LP setpoints.
The reduction in DNBR due to rod bowing is offset by a credit for low radial peaking in the critical assemblies.
Fewer than 73 fuel assem-blies will exceed the NRC-determined penalty threshold burnup of 24,000 MWD /T with a maximum burnup of 37,100 MWD /T.
The corresponding DNB penalty is 4.4 percent.
The power distributions for Cycle 2 show a maxi-N l
i
l
. mum radial peak for any of the 73 assemblies to be at least 15 percent Thus, the penalty is offset by the less than the maximum radial peak.
lower peaking of these assemblies and no power penalty for rod bowing is required for Cycle 3.
I The thermal hydraulic effects of sleeving and of the demonstration fuel assem-blies with reduced flow were found to be negligible in our Reference 7 SE.
{
2.2 Peaking Factor Uncertainties l;
In-core detector measurements are used to compute the core peaking factors using the INCA Code (Ref. h ).
The coef ficients required to perform this data reduction are performed using the methodology described in the Reference h topical report.
For Cycle 3 operation, NMEC0 has proposed measurement uncertainties of 6 percent for the total integrated radial peaking factor (Fr) and 7 percent for the total power peaking factor (Fq).
i
}
The initial CE evaluation of peaking factor uncertainty was presented i
in CENPD-145 and CENPD-153 (Refs h,1).
In a meeting with CE on March 6,1979, data was presented showing measurement uncertainty of f
6 percent in Fr and 7 percent in Fq to be conservative (Ref. 22).
On this basis, we find these measurement uncertainties to be acceptable.
j t
2.3 Burnable Poison Pin Performance I
l The burnable poison pins (also called " shim pins") in early CE Examination designed reactor cores have experienced perforations.
of the fuel discharged af ter Cycle I showed this problem.to be l
present at Millstone-2.
Therefore, it is necessary to examine the effect of continuing perforation in later cycles. The following table shows the history of shim pins' approximate exposure for the first i
four cycles of operation.
Number of Total Approximate l
Assemblies Number Shim Exposure l
Containing of at 80C Cycle Shims Shims (MWD /MTU) 1 108 1,296 0
t l
2 105 1,260 15,000 3
36 396 25,000 4
0*
0*
35,000
- Assuning no shims are introduced (none are planned) in the Cycle 4 fresh fuel.
2131 136
r t '
t In subsequent cycles af ter Cycle 1, the shim pins are relatively inert due to their high burnup of the contained neutron poison.
For Cycle 2 the shim perforation was analyzed by NNECO and we approved this analysis and judged the predicted consequences to be acceptable for one cycle of operation (Ref. 4).. Because of the X
much smaller number of shims and more inert nature of these shims in Cycle 3, we judge the Cycle 2 analysis to be bounding and further analysis not to be required.
Since normally shims are used only in first cycle fuel, and not in fresh reload fuel, we find shim perforation will not be a problem in Cycle 3 or in future cycles unless shims are introduced (not presently planned) in the fresh
'N s
reload fuel assemblics.
\\
2.4 CEA Guide Tube Integrity N
As the result of a scheduled fuel assembly examination of the Millstone-2 core following Cycle 1 operation, NNEC0 and CE~ reported Other CE designed severe guide tube wear in some fu"
- semblies. facilities have observed the same guide tube wear a i
f the CEA tips are parked in the " full up" position for long periods of j
I reactor operation.
The interim fix, proposed by CE and the licensees involve Our guide tubes previously damaged or to be used in CEA po i
l such as Reference 4 (Millstone-2) and Reference 7 (Calvert Cliffs These SE's conclude that "the sleeved guide tubes Unit No. 2).
will perform their function of reducing guide tube stress Any long term assembly is satisfactory for at least one fuel cycle.
j effects of relaxations of the mechanical " bulge" joint, including the possibility of radiation-enhanced relaxation, will have to be evaluated on selected assemblies at the next rr';eling outage."
l e
i All licensees have agreed to provide a guide tube evaluation The MNEC0 plan (Ref. 15 ),
program plan before their next refueling.as modified to incorporate i
The first inspection of sleeved CEA guide tubes was performed at The results Millstone-2 during the current Cycle 3 refueling outaae.
of this inspection program are reported in References 35, 38 and 42.
NNECO reports that the results of the eddy-current testing (ECT)
I indicated no detectable wear in any of the 50 guide tube sleeves inspected (Ref. 35).
However, some under expansion of the sleeves was indicated by the signal voltage measurement.
Two of the fuel assemblies with the highest indicated under expansion were scheduled for sleeve pull l
test.
2131 137.
~
,. j Reference 38 presents the results of the sleeve pull test for a total of 35 sleeves. Of the 35, 34 showed no axial movement at a pull force equal to the value used to prove intearity when the sleeves were first installed in 1978.
The remaining sleev.e moved 0.25 inches due to accidental application of approximately twice the normal test force.
During a subsequent pull test where sleeves in discharged fuel assemblies
.were to be pulled to destruct on (sleeve removal) to measure the force i
l required to remove the sleeves, one sleeve moved 0.25 inches at a force 4 lbs. below the nomal test force (Ref. 42). 'tiowever, application I
of a pulling force 1.5 times' the normal test force produced no further i
movernent. The pull-to-sleeve removal test showed that a minimum force of at I
least four times the nomal test force was required to remove any sleeve.
q We find that the ECT and sleeve pull test perfomed on sleeved fuel 4
assemblies after one cycle of operation under CEAs at Millstone-2 confirms the earlier out-of-core test and calculations showing that I
sleeving of CEA guide tubes provides an accepted interim repair to l
alleviate the guide tube wear problem for Cycle 3 operation, i
For Cycle 3 operation of Millstone-2, all but four fuel assemblies in CEA positions will be sleeved. To achieve this and to have sleeved fuel assemblies available for future cycles, 68 of the 72 batch E l
fuel assemblies will be sleeved prior to being placed in the core.
i The remaining four batch 3 fuel assemblies have been modified by decreasing the number and size of the flow holes and the size of the bleed holes.
Tests have indicated that the resulting decrease in guide tube flow was accompanied by less CEA flow-induced vibration and, therefore, less guide tube wear.
These four demon-stration fuel assemblies will be positioned in CEA Bank 7 locations.
Bank 7 is the last bank of CEA's to be pulled from the core and is I
used for reactivity control and some power shaping.
SE's for i
other CE facilities, such as Reference 7, have found a demonstration test similar to that proposed for Millstone-2, but with up to 16 i
fuel assemblies, to be acceptable.
The increase in the CEA insertion time to slightly more than 3.0 seconds (3.1 seconds for Millstone-2) was also found acceptable. We, therefore, conclude that the demon-stration test of four modified fuel assemblies with reduced guide l
tube flow is acceptable for. Cycle 3 operation of Millstone-2.
i l
NNEC0 has agreed to provide a Cycle 4 guide tube evaluation program, l
identifying changes from the Cycle 3 program at least 90 days prior to the Millstone-? shutdown for the Cycle 4 relodd outage.
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2131 138
n 14 -
7 t
2.5 Plant Systems Capability and Modifications As indicated in Section 1.0 of thisSE, major systems and components were designed for operation at 2700 MWt.
NNEC0 has performea a review that indeed verifies the correctness of the FSAR statement, that in fact, the major systems, structures and components are l
capable of supporting plant operation at the higher power level with little or no modification (Ref.14).
In References 8 and 11, NNECO identified their intention to make a system modification and to qualify a system as necessary to reach 2700 MWt.
These changes would be:
f l
e taking credit for charging pump flow in the small break LOCA analysis; and e adding a RCP speed sensing system (RCPSSS) RPS trip for pro-I tection against the four pump loss of flow incident.
The evaluations of these proposed modifications are presented in this section of the SE.
i 2.5.1 LOCA Credit for Charging Pump Flow l
NNEC0 proposed taking credit for charging pump flow in the small break LOCA analysis for increased power (2700 fMt) operation for l
Cycle 3 (Refs. 8,16 & 17).
A portion of the Chemical and Volume Control System (CVCS), although presently not credited in l
small break LOCA calculations, was originally designed, purchased and installed as a QA and Seismic Category 1 system. The QA and seismic requirements imposed on the system components are identical l
to those imposed on the remainder of the ECCS.
Reference 17 addresses the qualification of this systen.
i The portion of the CVCS presently utilized to inject boric acid I
into the core during emergency reactivity contrbi includes:
the two boric acid tanks; the two boric acid pumps; the three charging i
pumps, the regenerative heat exchanger; and the interconnecting piping and associated valves.
It is this same portion of the charging system that the licensee proposes to take credit for in small break LOCA analysis.
The charging pumps, during emergency operations, inject concen-trated boric acid into the reactor coolant system.
A pressurizer low level control signal or a Safety Injection Actuation Signal (SIAS) will automatically start all three charging pumps (only one pump is normally running).
The SIAS will also function to f
transfer the charging pump suction ~ from the volume control tank to the discharge of the boric acid pumps.
In the event that the boric acid pumps do not perform their intende(! function, an additional line has been provided to supply boric acid flow by gravity from the concentrated boric acid tanks to the charging pumps' suction header.
Portions of the CVCS, the charging pumps, the concentrated boric acid tanks, the boric acid' pumps, and associated valving and piping, have been designed to Seismic Category I requirements.
[N
The Millstone-2 FSAR and our SER have been reviewed to ensure that the charging system did initially meet the overall design requirements of General Design Criteria.
This review has concluded that the entire charging system has been designed to high quality standards in accordance with recognized codes, protected against natural phenonena, including a seismic event (except for the boric acid storage tank heaters as described below), designed to minimize the effects of fires, explosions, etc., designed to withstand the most severe postulated environmental conditions, and is not shared with Millstone-l.
The Millstone-2 SER further verifies the design adequacy and intent of the charging system.
SER Section 9.4.3 states, in part, that:
"The Chemical and Volume Control System (CVCS) is designed to..1) inject concentrated boric acid into the reactor coolant system during pressurizer low pressure, and/or a containment high-pressure signal.
Accordingly, a portion of the CVCS will be used as part of the emergency core cooling system in the event of a LOCA and is designed to Seismic Category 1 requirements."
2.5.1.1 Hechanical System The piping systems and mechanical components were found to be acceptable in performing the required mechanical functions properly.
Moveover, there have not been any new developments in NRC regulations or standards since that original review that would require any re-visions in the SER position regarding the mechanical operation of the system.
We conclude that the mechanical systems aspects of that portion of the charging system under review are acceptable.
2.5.1.2 ' Ele,ctrical Distribution c
The electrical equipment of the charging system required during emergency core cooling conditions derives its power from the station emergency AC and.DC power systems, Facility Z-1 and Facility Z-2.
These power systems (Z-1 and Z-2) are electrically and physically separate and redundant.
The licensing SER Section 8.3 concludes that the onsite power system satisfies GDC 17 and 18, IEEE-308 and IEEE-344, Regulatory Gyides 1.6,1.9 and 1.22 and is, thercfore, acceptable.
We have reviewed the Class IE distribution system for the charging system and determined that the system retains its original redundancy and electrical independency. We, therefore, find the AC and DC power systems to be acceptable.
4
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2.5.1.3 0_pgational Redundancy Upon initiation of SIAS, the three charging pumps receive a start signal.
If all three pumps are in service, two will start on one electrical bus and one f rom the other bus.
The swing pump (MP-18B) can be sunplied from Facility Z-1 or (-2.
Two redundant f'owpaths are provided.
First a gr.'v ity' feed flow-path from either boric acid storage tank through notor-0perated Valves (MOV's) 2-CH-508 or 509 to the common suction header for the charging pumps.
These parallel valves obtain their power from MCC-22-IE (Facility Z-1),
The second redundant flowpath utilizes the normal operating flow-path including separate parallel, suction lines from the boric acid storage tanks through the boric acid pumps P-19A and P-198.
Flow from the discharge of the pumps is directed to the suction l
header for the charging pumps via o single special line through isc ation valve M0V2-CH-514.
This valve plus the valves to isolate j
the boric dcid pump recirculation (Solenoid Operated Valves (S0V's) 2-CH-510 and 511) receive open and close signals upon SIAS, respectively.
All of this equipment in the second redundant flow-path is powered from Facility Z-2; the boric acid pumps and MOV2-j CH-514 from MCC-22-lF and the 50V's from 125 VDC panel C02R.
l l
The volume control tank isolation valve (MOV2-CH-SW ) and the isolation valve for the normal makeup line tc this tank (50V2-l CH-512), receiving close signals upon SIAS, are powered from Facility Z-1 powei sources.
Redundant charging line distribution i
50V's 2-CH-518 and 2-CH-519 are supplied 125V DC from Facilities I
Z-1 and Z-2, respectively.
Given a SIAS and fa' lure of either diesel generator th6 worst case sinole failure in the Design Basis Accident (DBA) analysis, the minimum f
quana' of boric acid that will be delivered from the boric acid stor-age taa. through the charging loops to the reactor coolant system, wil1 be the capacity of one charging pump.
NNEC0 proposes to take credit for 20 gpm, tbout one-half the flow of one charging pump.
1 Our analysis indicates that this portion of the charging system to be utilized in small break analysis has sufficient component re-i dundancy so that it is capable of withstanding the effects of a l
single failure.
Therefore, we find the operational redundancy acceptable.
2,f.1.4 Fire Protection Early warning fire detection will be provided in the charging pump area to give prompt warning of fire in the area.
Automatic sprinklers or suitable fire barriers will be provided to assure that fire damage does not result in a loss of shutdown capability where prompt action is not taken to suppress fires in these areas.
The consequences of fire damage to systems required for safe shut-down will be determined where the physical separation of cables may not preclude damage to redundant safety related systems. An ongoing study of separation for fire protection systems is in process and vill be cddrccsed in a supplement to the fire protection SER for
'Mll ":cra-: to be iss a: in the neaf futcre.
2131 141
t 2.5.1.5 Floodina It was determined that the portion of the enarging system being considered for small line break analysis is not subject to failure from flooding from non-category 1 (seismic) piping such as cir-culating water line and fire main.
We find this acceptable.
l 2.5.1.6 Quality Assurance _
The charging pumps were originally designed, procured and installed l
under the same NNECO and Bechtel quality assurance programs that i
were applied to other safety-related systems, including the High Pressure Safety Injection system.
In addition, these systems have been maintained in accordance with the NilEC0 quality assurance program for operations.
We, therefore, conclude that adequate QA require-ments have been established and applied to the charging pumps for their use as ECCS pumps.
l 2.5.1.7 Seismic Design The sei smic design requirements, including valve operability re-quirements, used for the design of the charging system are identical to those used on the remainder of the ECCS, except the documentation for seismic qualification of the boric acid storage tank heaters.
The licensee has documented that the heaters are not required to mitigate the consequences of the small break LOCA.
This contention I
will be evaluated in Section 2.8.3 of this SE.
The remainder of the charging system neets the seismic design requirements as specified in the Millstone-2 SER and is, therefore, acceptable.
i 2.5.1.8 Eny'ironmental Qualifications lI The portions of the charging system inside the containment have been -
I reviewed in regard to environmental qualification requirements, This review has indicated the acceptability of all components except i
solenoid operated valves CH-518 and CH-519.
These valves are re-l dundant charging line distribution valves remotely operated from j
the control room.
They do not receive a SIAS but must be open to l
allow charging flow into the RCS.
NilECO has reviewed the opera-tional requirements of CH-518 and CH-519 and determined that these valves are not required to operate during or following an accident.
They proposed that these valves be maintained open under administra-tive control (locked open) when the systen is required to be operable.
This is required by proposed TS 4.5.2.
Since CH-518 and CH-519 will be passive when locked open, they meet our environmental qualifications.
We, therefore, find the charging pump portion of the CVCS environmentally qualified.
I 2131 142
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i 2.5.1.9 Inservice Inspection and' Testing In accordance with the requirements of 10 CFR 50.55a, t1NEC0 has submitted proposed Inservice Inspection (ISI) and In-service Testing (IST) Programs for Millstone-2 to update the program to meet the requirements of the 1974 Edition through Summer 1975 Addenda of the ASME Section XI Code.
The program is currently under review by the flRC staff and will be addressed in a subseauent evaluation.
We have. made a preliminary review of flNEC0's ISI/IST submittals and found that the charging pump and related equipment will be inspected at the same frequency and in the same manner as other ECCS systems.
We find proposed surveillance requirements, TS j
4.5.2, to be consistent with TS requirements for other ECCS
- systems, t
l 2.E.1.10 Response Time I
NNECO has proposed that the charging pump response time be specified in TS Table 3.3-5, as less than or equal to 40.0 seconds.
This time i
allows 20 seconds for the diesel gen.rator to reach normal speed and voltage, 8.4 seconds for Sequencer Step 2 to close the charging pump breaker and the remaining 11.6 seconds for the charging pump te reach normal speed and discharge flow.
TS 4.3.1.1.3 requires that the response time of each reactor trip be demonstrated to be within its limit once per 18 months.
This response time will be an input parameter in the small break LOCA analysis.
We find the 40 second l
charging pump response time is acceptable.
l 2.5.1.11 j
Conclusion on LOCA Credit for Charging Pump Flow Based on our review of the licensee's submittals, the Millstone-2 l
FSAR, and our May 10, 1974 SER, we conclude that the portion of the charging system identified by the licensee as required for safety (small break LOCA) was qualified as an ECCS during the initial I
licensing review and continues to be qualified as an ECCS.
l l
We, therefore, conclude that the charging system is acceptable, I
for a minimum flow of half the capacity of one charging pump (20 gpm) in the j
ECCS Appendix K calculations for small break LOCA analysis.
i 2.5.2 RCP Speed Sensing RPS Trio NNEC0 proposed a modification to the RPS in conjunction with the request for ar. increase in the licensed power level (Refs. 9 & 21).
This modification involves installing a RCP Speed Sensing System (SSS) on the four reactor coolant pu ps which will provide a trip function to the RPS.
This trip function will provide a rapid and j
reliable reactor trip for the Four Pump Loss of Flow Event.
This trip function is presently being accomplished by the exi-ting steam generator differential pressure systen which protects the reactor againstlow reactor coolant flow due to the loss 'of one or More RCP's.
The stean generator differential pressure system will continue to 2131 143
~
. protect the reactor from low flow resulting from loss of less than four RCP's after the installation of the RCPSSS.
The four pump loss of flow has been identified as the most limiting anticipated transient in terms of requiring overpower protection.
The proposed RCPSSS, monitoring pump shaft speed,provi, des an anticipatory trip permitting a more rapid and accurate detection of the Four Pump Loss of Flow Event.
This allows a higher trip setpoint with no increased risk of sp,urious reactor trips.
e The RCP shaft sper' trip system consists o, a RCP Shaft SSS feeding f
into the existin APS trip logic and reactor trip switchgear.
The RCPSSS consists a speed sensor and mounting fixture, signal transmitter locat 3 inside containment, signal processing equipment, and analog bistable trip units for each RCP.
The analog bistable trip units, RPS trip logic and reactor trip switchgear remain as previously presented in the FSAR and evaluated in our May 10, 1974 SER.
Consequently, only the RCPSSS will be addressed in tnis evaluation.
The speed of each RCP motor is measured to provide a basis for the determination of a flow condition requiring reactor trip.
A metal disc with 44 uniformly spaced holes about its periphery and attached to the RCP shaft is scanned by a proximity device.
Each scanning device (proximity probe and transmitter) produces a voltage pulse signal with a frequency proportional to pump speed.
Signal proces-sing equipment modifies the pulse train signal from the scanning devices and provides input to a frequency to voltage converter which generates an analog voltage proportional to pump speed.
This analog voltage is compared to a fixed trip setpoint in the bistable trip unit.
A trip signal which trips the reactor is generated whenever the analog speed signal reaches or goes below the low trip setpoint.
The signal from the sensing probe is fed to its pulse transmitter located inside containment; a second probe and transmitter is provided for each pump.
Cabling from the pulse transmitter passes out of the containment and connects to signal processing equipment and a frequency to voltage converter lotatdd in its channel cabinet (four in total) in the control room.
Thq signal, after electronic processing, is fed through an analog bistable trip unit in the RPS cabinets to one of the RPS two-out-of-four logic channels.
All power and signal cabling design is in accordance with ecquire-ments for cable separation in Section 8.7 of the FSAR.
Cable current carrying capacities and tray / conduit fill are within acceptable limits in all cases.
Existing cable separation criteria are observed for the four independent signal channels and for the fcur separate N
120V at vital power source feeds (Zl, Z2, Z3 and Z4) supplying the SSS equipment.
The RCPSSS satisfies the. requirements of IEEE 279-1971.
We find this to be acceptable.
2\\5\\ \\kk
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.s s I' The RCPSSS consisting of the speed probe, pulse transmitter, signal processor, and speed probe extension cable is functionally identical to the corresponding system provided as part of the core protection calculator system in the RPS for Arkansas Nuclear One, Unit 2 (ANO-2),
i Docket No. 50'368.
This equipment was envirormentally qualified in accordance with IEEE 323-1971 to the following specifications:
total Gama dose 1.2 E7 RADS, 212 F at 100 percent RH with chemical spray for equipment located inside containment.
The sicnal processing assem-bly whi.ch is located in the control room is qualified for background l
radiation,40-135 F and 20 - 95 percent RH.
The test report dated September 30, 1976 was submitted, during the licensing process, as part of the FSAR for the staff's review. The staff concluded in its licensing SER for ANO-2 that this equipment is environmentally qualified.
j NNEC0 has documented that operation of the RCPSSS is not required during or subsequent to any Design Basis Event which significantly alters the contaiament environment (LOCA, Main Steam Line Break, Feedwater Line Break); therefore, it is not required that in-containment equipment be qualified for the adverse environments I
associated with these events.
l The Devar frequency to voltage converter model 18-116 located in i
the control room is different than that supplied for the ANO-2 project.
The converter was purchased and qualified to the following environmental parameters:
temperature 60-137 F, humidity 40 -
95 percent pressure atmospheric, and radiation negligible.
These environmental considerations equate to control room expected environ-ment, and therefore, are acceptable.
I The licensee has further stated that the equipment has been procured, l
tested and qualified to IEEE 323-1971. We conclude that the RCPSSS equipment as described above and tested in accordance with IEEE 323-1971 is environmentally qualified.
i l
The seismic test information for the RCPSS system components, proximity probe, conne'ctor, pulse transmitter and signal processor was submitted in CENPD - 182, " Seismic Test Documentation," data sheet 2j and 205.
WP'le our review of CENPD-182 as a generic testing documentation is not completed, NNEC0 submitted additional data in Reference 33 to demonstrate that the Test Response Spectrum (TRS) enveloped the Required Response Spectrum (RRS) for the above RCPSS system components of Millstone-2.
The seismic test documentation for the Devar frequency to voltage converter was also documented in Reference 33.
{
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213 145
' For each of the four primary components of the speed sensing system,
[
namely the proximity probes, transmitters, signal processors, and frequency-to-voltage converters, the corresponding TRS completely envelopes the Plant Mounting RRS of Millstone-2.
The transmitters and proximity probes were tested using sine beat tests.
The signal processor and frequency-to-voltage converter were tested biaxially using multi-frequency random inputs in accordance with IEEE 344-1975.
It is our conclusion that the seismic qualification test of the signal processors and frequency-to-voltage converters of the RCPSSS is acceptable for use in Millstone-2.
The seismic qualification of the proximity probes and transmitters of the same system using sine beat testing methods is acceptable for an interim period of one fuel cycle.
It will be required that the proximity probe and transmitter be retested using a multiple-frequency and multiaxis test in accordance i
with IEEE 344-1975 prior to the startup from the Cycle 4 refueling outage.
NNEC0 proposed TS modifications to include this new RPS trip in the RPS instrumentation trip setpoint limit table (Table 2.2-1) and operability, response time and surveillance requirements.
We considered an additional T5 to confirm the RCPSSS flow signal with the steam generator delta-pressure flow signal.
However, existing TS 4.2.6 requires the RCS flow rate to be determined greater than 370,000 gpm at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> to meet DNB margin requirements.
If the RCS flow decreases below the limit, a reactor shutdown is required by TS 3.2.6.
Based on our review of the licensee's subnittals, we conclude that the modifications to the reactor protection system satisfy the require-ments for single failure, electrical isolation and physical separation, and environmental qualifications, and therefore, are acceptable.
It is further concluded that the TS as proposed are acceptable.
We therefore, conclude that the installation of the RCPSSS to initiate a trip of the reactor upon loss of four reactor coolant pumps is acceptable.
2\\51 146
2.6 Analyses of Anticipated Operational Occurrences ( A00s)
References 18 and 20 present the safety analysis of postulated A00s for f tillstone-2.
NNECO classifies the itst of postulated A00s into l
two categories.
Anticipated Operational Occurrences for which the RPS Referen.ce Analysis Assures no Violation of SAFDLs:
Analysis Status Control Element Assembly Withdrawal Cycle 2 Reanalyzed Reactor Coolant System Depressurization Cycle 2' Reanalyzed Boron Dilution Cycle 2 Reanalyzed l
Loss of Load FSAR Reanalyzed Loss of Feedwater Flow FSAR Reanalyzed f
Excess Load FSAR Raanalyzed Startup of an Inactive Reactor Coolant Pump FSAR Not Analyzed i
Excess Heat Removal due to Feedwater Malfunction FSAR Not Analyzed Anticipated Operation 11 Occurrences which are Dependent on Initial Overpower Margin for Protection Against Violation of SAFDLs:
Loss of Coolant Flow Cycle 2 Reanalyzed Full Length CEA Drop Cycle 2 Reahalyzed j
Transients Resulting from Malfunction of One Steam Cycle 2 Reanalyzed Generator l
l l
t I
t k
1
, The first category includes those A00s for which RPS trip setpoints assure that the Specified Acceptable Fuel Design Limits (SAFDLs) are not exceeded.
Protection is provided by a reactor trip.
The second category includes those A00s for which initial steady-state overpower margins are maintained by LCOs to assure that the SAFDLs are not exceeded.
Protection is provided by the margin and a trip may not be required.
The LCOs and LSSSs for Cycle 3 TS were calculated using the methods described in Reference r.
This CE topical report is currently unde.r review by the NRC staff.
The review h;s progressed sufficiently to conclude that its application for this purpose is acceptable.
The transient plant responses used in A00 reanalyses were calculated using the computer code CESEC (Ref. c).
CESEC models both the primary and secondary coolant loops and accounts for the effects of all equipment response in both loops.
CESEC is currently under review by the NRC staff.
This review has progressed to the point that there is reasonable assurance that the results dependent on CESEC will not be appreciably altered by any revision resulting from our review.
NNEC0 stated that the need for reanalysis of a particular A00 is deter-l mined by comparisen of the key parameters for that A00 to those of the l,
last cycle for which a complete analysis was performed.
If the key parameters are within the envelope of the Reference Cycle data, no reanalysis is required.
A reanalysis might also be performed if it leads to a significant relaxation of TS restrictions.
The transients not reanalyzed and the justification for omitting reanalysis is:
- Excess Heat Removal Due to Feedwater Malfunctions - power i
increase reduces the core reactivity effect since the available excess feedwater flow is reduced; event bounded by 2560 MWt analysis in FSAR.
part length CEA Events - the part length CEA's have been removed (Ref. 4).
We find that the 1;censee has ranalyzed the appropriate A00s for Cycle 3 operation at 2700 MWt.
Those transients which were reanalyzed are discussed below.
In all cases our acceptance criteria are the same as the criteria used to judge previous cycle analyses.
1 l
2131 148
F-.
. i l
2.6.1 Safety Analysis Parameter Values l
The major core parameters assumed in the safety analysis for Cycle 3 are the following:
l REFERENCE CYCLE 2 CYCLE 3 l
i PARAMETER UNITS VALUES VALUES i
l Power Level MWT 2611 2754 I
f Maximum Steady State Core Inlet Temp F
544 551 f
Minimum Steady State RCS Pressure psia 2200 2200 I
Maximum Steady State RCS Pressure psia 2300 2300 l
i Minimum Reactor Coolant Core Flow E6 lb/hr 134.9 133.7 (2200 psia, 551 F)
ASI DNBR LCO Limit at Maximum ASI Units
.21
.16 Allowed Power Level Maximum CEA Insertion at Maximum
% Insertion of 25 25 Allowed Power (2700 MWt)
Group 7 Maximum Allowed Initial Peak kw/ft 16.0 16.0 Linear Heat Rate (DBEs Other l
than LOCA) l Steady State Linear Heat kw/ft 21.0 21.0 Rate to Fuel Centerline i
Melt l
CEA Worth at Trip - HFP*
E-2 ap Steam Line Break
-5.25
-5.31 Steam Generator Tube Rupture
-5.41
-4.32 CEA Withdrawal
-4.60
-4.32 Loss of Load
-2.40
-3.20 l
CEA Worth at Trip - HZP*
E-2
-3.2
-3.2 30 Steam Line Break
-3.2
-3.1 CEA Withdrawal sec.
2.75 3.10 CEA Time to 90% Insertion
(.ncluding Holding Coil Delay)
RTD Response Time sec.
5.0 8.0 r
I I
2131 149
I REFERENCE PARAMETER UNITS CYCLE 2 CYCLE 3 l
VALUES VALUES Total Planar Radial Peaking Factors For DNB Margin Analyses (Fr)
Unrodded Region 1.440 1.598 Bank 7 Inserted 1.550 1.806 For kw/f t Limit Analyses (Fxy) j Unrodded Region 1.540 1.584 Bank 7 Inserted 1.660 1.822 l
Peak Augmentation Factor 1.045 1.054 Moderator Temperature Coefficient **
E-4 ar/F
-2.5 to
-2.5 to
+.5
+.5 l
l Shutdown Margin (Value Used in Zero
% op
-3.2
-3.2 i
Power SLB)
I i
i i
- Note: Different values are used for different analyses in order that current analyses may be bounded for future cycles.
In all cases i
the assumed values must be demonstrated to be conservative.
methodology used for Cycle 3 operation of Millstone-2 at 2700 MWt initially l
computes a TM/LP setpoint which guarantees that the DNBR does not decrease below 1.19 for all steady state conditions.
This steady state TM/LP setpoint cannot be used directly because during a transient the DNBR l
degrades for a short time after the trip setpoint is reached.
To account i
for this, a Time Adjustment Bias (TAB) is added to the steady state TM/LP setpoint so that the resultant dynamic TM/LP setpoint guarantees a trip l
soon enough that in the course of the transient, core conditions will not degrade beyond the steady state TM/LP setpoint.
2131 150 i
=
l
. The TAB is determined by the transient for which DNBR suffers the greatest degradation after the trip setpoint is reached.
All potentially limiting transients must be considered in computing the TAB.
The Loss of Flow, Seized Rotor, CEA Drop, CEA Ejection, Steam Line Rupture, and Malfunction of one Steam Generator events do not require a TM/LP trip, and hence are excluded from consideration in the determination of the TAB.
For a transient to be limiting in the determination of the TAB, it must produce a rapid degradation in DNBR.
The rate of DNBR degradation was assessed for all transients, and it was determined that the CEA Withdrawal Event and RCS Depres-surization Event were limiting by a wide margin.
Thus only these two events were considered in the computation of the TAB.
The TAB determined by the CEA Withdrawal and the RCS Depressurization events were 45 psiPand 30 psi respectively.
The 45 psi TAB oetermined by the CEA withdrawal is limiting, and hence was used to adjust the TM/LP setpoint.
The analysis of these two events in Reference 18 assumes a reactor trip at the point of greatest rate of degradation of DNBR, and not at any realistic trip time.
For the purposes of computing the TAB, this is a conservative approach.
Because of the arbitrary trip time, the minimum DNBR quoted for these two events in the reload application has no safety significance.
The dynamic TM/LP trip, which includes the TAB adjustment, guarantees a minimum DNBR of 1.19 for the most limiting CEA Withdrawal Event and approximately 1.21 for the RCS Depressurization Event.
For all other DNBR events the minimum DNBR quoted is that which would occur assuming the TM/LP setpoint computed as described above is in effect.
2.6.3 Computation of Rcquired Overpower Margin from Transient Analyses The Required Overpower Margin (ROPM) is a measure of the maximum DNBR degradation from the LC0 which occurs in the course of any transient.
The limiting transients are those which either require no reactor trip (CEA drop and malfunction of one steam generator) or else require a low RCS flow trip (two pump loss of flow and four pump loss of flow).
All four transients are analyzed parametrically in burnup, power, CEA insertion, ASI and axial power shapes consistent with these four para-meters.
For a given transient and any given power and ASI the computation with the combination of other parameters which maximize the R0PM gives the R0Pf' value.
For each value of ASI the maximum R0PM determined from the four transients is chosen and this R0PM is used to adjust a DNBR LC0 determined for static conditions to provide a DNBR LC0 applicable for dynamic conditions, which is the DNBR LCO used in the TS.
This dynamic DNBR LC0 guarantees that during the course of any of these four transients the DNBR will not decrease below 1.19.
NUEC0 states that for this reload the Four Pump Loss of Flow was the limiting transient for all values of the parameters considered.
Thus the R0Pfi from the Four Pump Loss of Flow was used to compete the whole of tho D';SR LCO.
This being the case, a Four Pump Loss of Flow event initiated fron any point on the DNBR LC0 will result in a minimum DNCR of 1.19.
The analysis presented for this reload is one such case, and as exnected the minirm DNBR is 1.19.
We find this methodology McepNble.
2131 151
In keeping with the previous discussion, in the analysis of the CEA Drop Event (Section 2.6.10) and the Malfunction of One Steam Generator event (Secticn 2.6.11), the most deleterious combination of initial conditions determined from the parametric study were used in the cases presented in the reload application.
Hence the minimum DNBR given for I
these two events is the minimum which can occur starting the event
}
from any initial condition consistent with the DNBR LCO.
2.6.4 CEA Withdrawal l
l The CEA Withdrawal Event required reanalysis because the CEA drop time has increased tc 3.1 seconds (See Section 2.4) from the 2.75 second value used in the Reference Cycle analysis. The CEA worth has also decreased from the value used in the Reference Cycle t
i analysis.
As a result of these changes the TM/LP setpoint TAB com-1 puted for the CEA Withdrawal Event has increased from 30 psi in the Reference Cycle analysis to 45 psi in the present analysis.
This TAB of 45 psi is reflected in TS Section 2 changes.
We find this analysis and the TS changes acceptable.
The CEA Withdrawal Event from flZP was also reanalyzed.
In this case the reactor trips on the Variable High Power Trip at 25% power, l
rather than the TM/LP trip.
The core power momentarily spikes to 141% of 2700 MWt, but the heat flux reaches only about 61.5% of 2700 MWt and the minimum DNBR reached is 1.58.
We find this analysis and its consequences acceptable.
t i
2.6.5 RCS Depressurization l
The RCS Depressurization Event was reanalyzed for Cycle 3 because the CEA drop time has increased and the CEA reactivity worth ha., decreased from the Reference Cycle values. The CEA Withdrawal Event is one of those analyzed l
to determine the limiting time delay adjustment to y in the TM/LP trip setpoint.
The RCS depressurization time delay bias adjustment was computed to be 35 psi, which is less limiting than the CEA Withdrawal Event value of 45 psi. Hence, the adjustment used will be 45 psi. We find this analysis to be acceptable.
2.6.6 Boron Dilution i
l The Boron Dilution Event was reanalyzed because tne critical boron con-centrations have increased from their Reference Cycle analysis values.
i NRC criteria require that, given a Boron Dilution Event, time is pro-vided for the operator to prevent a return to a critical condition of at least 15 minutes for the startup, hot standby, hot shutdown, and cold shutdown modes and a minimum of 30 minutes for the refueling mode.
The limiting dilution A00s for Millstone-2 Cycle 3 are for the cold shutdown mode with a time to criticality of 20 minutes and for the re-fueling mode with a time to criticality of 44 minutes.
Thus, the times calculated for Cycle 3 meet the NRC criteria, and we find this analysis and its results acceptable.
2131 152
' j While the NRC criteria require that it be demonstrated that sufficient time is available for operator intervention during a Boron Dilution Event when the reactor is critical, no credit is given for operator intervention, and the analysis must demonstrate that the consequences of a Boron Dilution Cvent 'd thout operator intervention are acceptable. Without operator ut'rvention a Boron Dilution Event from power operation is terminated by the variable high power trip, i
the local power density trip, or the TM/LP trip.
NNEC0 states that the most severe Boron Dilution Event would be less severe than the l
CEA Withdrawal Event because of the significantly slower reactivity insertion rate in the Boron Dilution Event.
Hence, the CEA Withdrawal Event analysis bounds the Boron Dilution Event analysis and a separate analysis is not required for the Baron Dilution Event. We concur with their statement.
l 2.67 Loss of Load l
The loss of Load Event was reanalyzed for Cycle 3 because the power level, core inlet temperature and CEA insertion time have increased and the RCS flow and pressure have decreased from the Reference Cycle values. With' the exception of RCS pressure all these changes represent l
actual changes in Cycle 3 parameters.
The kCS pressure used was changed from the highest value allowed by the TS to the lowest value allowed by the TS because it was found that the Loss of Load Event is more severe if the initial RCS pressure is low. The Loss of Load Event results in:
- 1) a minimum DNBR of 1.33 which is well above the limit of 1.19 and 2) a maximum RCS pressure of 2555 psi which is well below the upset pressure limit of 2750 psia. We find this analysis and its consequences to be acceptable.
2.6.8 Loss of Feedwater Flow The Loss of Feedwater Flow Event was reanalyzed because the core power, radial peaking factor, core inlet temperature and steam generator pressure have increased from their Reference Cycle values and RCS mass flow has decreased from its Reference Cycle value.
The loss of feedwater flow results must meet two criteria.
First, the DNBR must not fall below 1.19 with total loss of feedwater. The analysis predicted a minimum DNBR of 1.33, which is well above the limit.
The DNBR consequences of this event are more severe if it is assumed thct the steam dump system, steam bypass sytem, and pressurizer relief valves are inoperable, and they were assumen inoperable in the DNBR analysis.
Thus, in this analysis all cooling is through pressurizer safety valves, steam safety valves, and CVCS.
Failure of a steam safety valve
~
i to reclose would result in ar A00 bounded by the Excess Load Event.
i The reactor is predicted to trip in 31 seconds via the high pressurizer pressure ;ignai.
I 2131 153 i
j
i Second, the analysis must ensure that following a Loss of Feedwater Flow Event, the steam generators do not go dry in less than 10 minutes following the event and the resultant reactor trip, thus allowing the operator at least 10 minutes to reinitiate normal or auxiliary feedwater flow. The analysis showed that after 10 minutes approximately 10 percent of the steam generator inventory remained, l
and that the steam generator would not go dry until 15 minutes after
}
the loss of feedwater flow. The drop in steam generator level is more
}
severe if it is assumed that the steam dump system and steam hypass system are operable, and they were assumed operable in this calysis.
We find these analyses and their consequences to be acceptable.
Failure of a pressurizer safety or relief valve to close after a +.otal Loss of Feedwater Flow Event, Threa Mile Island Unit No. 2 (THI-21 l
Accident, will be addressed in a separate safety evaluation.
2.6.9 Excess Load The Excess Load Event was reanalyzed because the core power, temperature, l
and radial peaking factors have increased and the RCS flow has decreased from the Reference Cycle values.
Rather than rqanalyzing a number of cases, as is usually done, the licensee first enmined the input parameters for computing the TM/LP from the Excess Load Event and found them to be conservative relative to the values used for 1
determining the TM/LP for the CEA Withdrawal Event. This demon-strated that in an Excess Load transient the TM/LP trip will guarantee I
that the DNBR does not,o below 1.19.
l l
Next, they performed the transient analysis on what was considered the i
worst case. The minimum DNBR predicted is 1.41.
We find this method I
of analysis and its results to be acceptable.
2.6.10 Loss of RCS F'.ow The Four Pump Loss of Flow Event was reanalyzed because the core power and inlet temperature have increased from the Reference Cycle analysis values and j
an RCP speed sensing system has been added to the RPS system (See Section 2.5.2 1,
of this SE).
The RCP loss of flow RPS trip performance characteristics are:
i i
PERFORMANCE CHARACTERISTICS LOW FLOW TRIP FUNCTION RCPSS TRIP FUNCTION (STEAM GENERATOR P)
{
b I
System Overall Accuracy 2.7%
1.5%
Bistable Drift Allowance 0.8%
0.8%
System Noise 2.25%
1.0; Nominal Trip Setpoint 91.7%
93%
Overall System Response Time 650 msec 450 msec 2131 154
NNEC0 states that this new system increases the RPS performance by allow-ing a higher RCP loss of flow trip without increasing the risk of spurious reactor trips (Ref. 9).
The combination of a higher trip setpoint and a more rapid system response assures a reactor trip sooner during this Four Pump Loss of Flow Event.
The trip setpoint, equivalent to 93 percent flow, I
is829 rpm (Ref.25).
t The change from a low flow trip on ap flow to a low flow trip on RCP speed requires a slight change in the analytical methodology for the Four Pump Loss of Flow Event. We have reviewed this change in methodology and find it reasonable.
Since we have accepted the pre'.)ous methodology l
for reloads, with this one change we find the present methodology to be l
acceptable.
NNEC0 has not presented the analysis of the Two Pump Loss of Flow Event, but has stated that it was analyzed and determined to be less limiting than the Four Pump Loss of Flow. NNEC0 also states that the Two Pump Loss of Flow Event is more limiting tnan the One Pump Loss of Flow l
Event, and hence the One Pump Loss of Flow Event does not require I
analysis. We find both these statements to be reasonable, and accept them without detailed analysis.
2.6.11 CEA Drop j
i The CEA Drop Event has been reanalyzed because the power level, radial peaking, and core inlet temperature have been increased from the Reference Cycle values and the core flow has decreased from the Reference Cycle l
value. The minimum DNBR attained is 1.21 which is slightly greater than our acceptance criterion of 1.19.
We find this analysis and result to be I
acceptable.
i 2.6.12 T.'ansients Resulting From Malfunction of One' Steam' Generator l
There are four events in this category, which are:
e Loss of Load to One Steam Generator i
e Excess Load to One Steam Generator i
e Loss of Feedwater to One Steam Generator l
e Excess Feedwater to One Steam Generator i
These four events are analyzed in Reference r.
The Loss of Load to j
One Steam Generator has been found to be the limiting of the four i
events by a wide margin.
Hence, it is only this event which needs to be l
analyzed.
The Loss of Load to One Steam Generator requires reanalycis because the power level, radial peaking, and core inlet temperature have I
increased and the core flow has decreased from the Reference Cycle values.
l The analysis predicts a minimum DNBR of 1.24 which is above our criterion of 1.19 and predicts a Linear Heat Rate (LHR) of 19 kw/f t which is below our centerline melt criterion of 21 kw/f t.
We find this analysis and its results to be acceptable.
2131 155
i- -
I j
2.7 Postulated Accidents Other Than LOCA t
l The following postulated accidents were reanalyzed for Cycle 3 to ensure that in their event the extent of fuel failure and subsequent radioactive release will be acceptable.
The radiological consequences are evaluated in Section 2.10.
2.7.1 Steam Generator Tube Rupt.ure The Steam Generator Tube Rupture Event was reanalyzed for Cycle 3 because the l
power level, inlet temperature and RCS pressure have increased and the CEA worth has decreased from the Reference Cycle values.
For this event, DNBR protection is provided by the TN/LP trip, and NNEC0 states that sufficient margin is available that no DNBR analysis of the transient is necessary.
We find this appraisal to be reasonable.
The Steam Generator Tube Rupture Event analysis is performed to determine the maximum site boundary dose.
The Cycle 3 analysis shows a greater leakage l
rate from the tube rupture, but assumes the steam generator is 1solated j
by operator action in 18 minutes rather than in 30 minutes as assumed in the FSAR analysis. As a result, the Cycle 3 analysis shows an integrated primary to secondary leakage of approximately 48,000 lbs. which is less than the FSAR-assumed leakage of 53,200 lbs. Thus, the FSAR analysis bounds this l
reanalysis in regard to primary to secondary leakage.
I j
2.7.2 tiain Steam Line Rupture '
The flain Steam Line Rupture Event was reanalyzed because the power level, core l
inlet temperature and steam generator pressure have increased and the boron j
worth has decreased from the Reference Cycle values.
Both the Hot Full Power (HFP) EOC and Hot Zero Power (HZP) E0C cases were analyzed, i
For the HFP case, the reactor remains subcritical by at least 0.15 percent op throughout the transient.
The power drops quickly to approximately 9 percent and there is a brief return to a power level of approximately 12 percent at 72 seconds, the time of steam generator dryout and lowest primary coolant temperature.
After this, the power level drops to shutdown conditions.
For the HZP case, the reactor remains subcritical by at least 0.15 percent op and thus there is no return to power.
For both the HFP and HZP cases, the minimum DNBR is greater than 1.19 and hence no fuel failure is predicted.
All radioactive release is from radioactive material already in the secondary coolant loop and from primary to secondary j
leakage.
This is the same situation as was analyzed in the FSAR, 2131 156
32 -
1 2.7.3 CEA E,{ection l
The CEA Ejection Event was reanalyzed because the pJwer level, the post-ejected peak power and augmentation factors have increased from the Reference Cycle values. Also, the ejected rod worth and delayed neutron fraction have decreased from the Reference Cycle values. Both the HFP and HZP cases were analyzed, and it was found that j
the HFP case was more limiting.
j The CEA Ejection Event is very rapid, and NNEC0 states that the total energy deposition is the proper parameter to monitor, rather than LHR or DNBR as is the case for other transients. The specific criteria employed are:
o Clad Damage Threshold Total Average Enthalpy=200 calories / gram Incipient Centerline Melting Threshold o
Total Centerline Enthalpy=250 calories / gram Fuily Molten Centerline Threshold e
Total Centerline Enthalpy=310 calories / gram The analysis performed by NNEC0 predicted a maximum total average l
enthalpy of 190.8 calories / gram and a maximum centerline enthalpy of 287.2 calories / gram with 2.1 percent of the fuel pins suffering incipient centerline melting.
Since there are no cladding failures, NNECO assumes no radioactive release from these fuel pins suffering centerline melting.
I We judge the NNECO analysis to be conservative and find their l
predicted consequences acceptable.
l 2.7.4 Seized Rotor The Seized Rotor Event was reanalyzed because the power level, core inlet temperature, and radial power peaking have increased and the core flow has decreased from the Reference Cycle values.
In the Seized Rotor Event, it is normal to predict a small fraction of the fuel will suffer DNB, and the purpose of the analysis is to assure that the radioactive release which would result from this event is acceptable. The Cycle 3 analysis of the Seized Rotor Event predicts 1 percent fuel failure through DNB. An NRC analysis for the Steam Line Break Event shows that it would be necessary to fail 4 percent of the fuel to exceed the 10 CFR 100 dose limit.
Since the Seized Rotor Event is predicted to fail only 25 percent this much fuel, the radioactive release could, at most, be approximately 25 percent of the 10 CFR 100 limit.
We find this analysis and its results' acceptable.
2131 157 f
iI.
I
l 2.8 Evaluation _of LOCA Analysis NNEC0 has provided reanalysis of the Large Break and Small Break LOCA analysis for Cycle 3 (Refs. 24, 30, 33).
Our review has shown a problem with the CE analysis of the rupture strain model used in LOCA calculations. This is evaluated in the following sections.
l i
2.8.1 Rupture Strain Model Review i
In order to comply with 10 CFR 50, Appendix K, the LOCA analysi_s_
must demonstrate that the peak clad temperature (PCT) remains below i
l 2,200 F and the maximum local cladding oxidation, which is a function of the time dependence of the PCT, remains below 17 percent.
During a LOCA, the cladding swells due to the decreased coolant pressure and the increased fuel temperature and gas pressure.
l The clad swelling is terminated if the cladding ruptures.
The clad swelling and rupturing affect the PCT through the following l
mechanisms:
Swelling causes flow blockage which increases PCT.
e Swelling causes greater cladding oxidation (metal-water reaction) e i
due to the greater surface area which increases PCT.
Swelling increases the gap conductance which decreases the PCT.
l e
Swelling increases the area for convective heat transfer which e
decreases the PCT.
Swelling increases the area for radiative heat transfer which decreases i
e I
the PCT.
t Clad rupture leads to metal-water reaction inside the cladding which l
e i
increases the PCT.
For rupture during the blowdown phase, as is predicted for Millstone-2,
~
these competing effects are nearly balanced, and it is not clear whether an upper bound or a lower bound to a rupture strain curve would be the more conservative assumption. To date, we have accepted a compromise position which utilizes a best estimate curve. This methodology is described in CENPD-136 (Ref. W ).
We have independently assessed the available data from experiments and constructed a rupture strain curve to compare with that used for Millstone-2. At the point where rupture was predicted'to occur our curve and the Millstone-2 curve are in good agreement (38 percent strain vs 32 percent strain). This difference is substantially less than the scatter in the data from experiments, f
On this basis, we find the strain rupture curve for Millstone-2 adequate for use in LOCA analyses.
2131 158 c
. 34 2.8.2 Large Break _LOCA NNEC0 submitted the large break LOCA analysis for Cycle 3 operation at 2700 MWt in References 30 and 32 The evaluation was performed at a reactor power level of 2754 MWt (10* percent of 2700 MWri and a peak c
linear heat generation rate (PLHGR) of 15.6 kw/f t.
A spectrum of break sizes was analyzed using NRC-approved large break ECCS evaluation methods (Ref. d ).
The double-ended cold leg guillotine rupture at the pump discharge (DEG/PD) with a discharge coefficient of 0.8 was found to be the most limiting break. The 0.8 double-ended slot break (DES /PD), which was the limiting break for Cycle 2, is almost equally limiting.
The most limiting fuel rod for Cycle 3 operation is a rod in one of the partially depleted batch 8 assemblies retained from Cycle 2.
For this rod, clad rupture is predicted to occur during the blowdown. Clad rupture during blowdown leads to high clad temperatures due to degradation of the cooling of the rod during blowdown and decreased effectiveness of rod-to-rod thermal radiation during reflood.
At the time of minimum fuel-clad gap conductance, the fuel pin pressure was not high enough to cause rupture during blowdown. Therefore, the highest clad temperatures were predicted at the time when fuel pin pressure.first-bec~ame high enough to cause a blowuown rupture,,,... -
The calculated peak clad temperature is 2081 F and the calculated maximum local metal / water reaction percentage is less than 16 percent. These are below the acceptance criteria specified in 10 CFR 50.46 and we, therefore, find this large break LOCA analysis acceptable.
2.8.3 Small Break LOCA NNEC0 submitted their analyses of a spectrum of small breaks at a PLHGR of 16.0 kw/f t in References 24 and 32. These analyses were performed with approved CE models for small break LOCA calculations (Ref, h).
As discussed in Section 2.5.1, the portion of the charging system required for small break LOCA safety is qualified as an emergency core cooling system, with the exception of the boric acid storage tank heaters (See below). The charging s;-tem, as qualified, is acceptable for a maximum flow equal to the capacity of one charging pump for ECCS Appendix K calculations. The licensee assumes that only slightly less than half of the flow of one pump (20 gpm) is delivered to the core due to the break location. This 20 gpm is combined with the high pressure safety injection (HPSI) flow rate in the analysis.
2131 159
The limiting small break was determined to be the 0.1 sq ft cold leg break l
in the pump discharge leg. The analysis for the last cycle identified the 0.05 sq ft as the limiting break. The CE small break evaluation model report demonstrated that the 0.1 sq f t break is more limiting than the 0.05 sq f t break (Ref. h ).
The NNECO analysis showed that for breaks smaller than 0.1 sq f t, core uncovery begins later when the decay heat generation is less, so the amount of fuel assemblies uncovered is less. Breaks larger than 0.1 sq ft are less limiting since the faster depressurization rate results in l
safety injection tank actuation.
l l
The limiting fuel performance occurs at the time in life of minimum gap conductance (highest stored energy) for a partially depleted B fuel assembly rod.
For the 0.1 sq f t break with 20 gpm charainq flow, the peak clad temperature l
and peak Zirconium oxidation percentages are 1971 F and 10.3 percent respectively.
The acceptance criteria limits are 2200 F and 17 percent.
The documentation of the seismic qualification of the Boric Acid Storage Tank (BAST) heaters is not presently available. No credit, then, can' be taken for I
these heaters in the small break LOCA analysis. During the time period in which the charging pumps are assumed to be operating, the tank temperature must remain high enough to prevent boron precipitation.
The peak clad temperature for the worst small break is reached in 24 minutes.
Calculations of tank heat losses show that the temperature does not drop below the minimum TS temperature limit for more than four and a half hours.
This result is conservatively based on a heat loss rate detennined for a full i
tank and a heat content applicable to the minimum tank volume.
Thus, seismic qualification of the BAST heaters is unnecessary because they are not needed in the time required to mitigate the consequences of a small break LOCA.
i I
l Therefore, we conclude that the use of 20 gpm charging flow in the small break LOCA analysis is acceptable, and that the LOCA analysis demonstrates confonnance i
with the Acceptance Criteria for ECCS for Light Water Nuclear Power Reactors (10 CFR 50.46).
l l
l i
i i
l l
i 2.9 Meteorological Data Review l
We have reviewed the meteorological information provided in Reference 14.
We calculate the short term atmospheric dispersion factors (X/Q) for j
both ground level and stack releases using the 1974-1975 data to be:
Ground Level Release Time Period Distance and Direction X/0 (sec/ cu m) 0 - 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> 600m SSW (EAB) 4.3 x 10~4 0 - 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 3940m SSW (LPZ) 2.4 x 10 -5 l
8 - 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 3940m SSW (LPZ) 1.6 x 10
~
-5 l - 4 day 3940m SSW (LPZ) 6.6 x 10 4 - 30 day 3940m SSW (LPZ) 1.8 x 10 -5 l
114 Meter Elevated Release I
i Time Period Distance and Direction X/Q (sec/cu m)
~*
0 - 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> 503m (EAB) 9.4 x 10 l
O - 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> 3860m (LPZ) 2.5 x 10 -g*
-6 f
4 - 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 3860m NNE (LPZ) 4.1 x 10
~0 8 - 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 3860m NNE (LPZ) 1.8 x 10 1 - 4 day 3860m NNE (L9Z) 7.8 x 10 -7
-7 h
4 - 30 day 3860m NNE (LPZ) 2.3 x 10 30 Meter Release Time Period Distance and Direction X/0 (sec/cu m)
~4 0 - 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> 620m (EAB) 5.5 x 10
~4 0 - 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 3940m (LPZ) 1.1 x 10
- Fumigation with wind speed = 2m/sec 2131
)61 s
l I Our calculations were made using the direction-dependent meteorology l
l model described in draf t Regulatory Guide 1.XXX, " Atmospheric Dispersion Models for Potential Accident Consequence Assessments at I
Nuclear Power Plants" (October 26,1978).
The X/Qs for accidents involving the secondary coolant system were calculated using the method in Regulatory Guide 1.5 for determining i
l X/Q values. This method assumes a 30m release height with fumigation, l
atmospheric stability class F, and wind speed of 1 m/s.
The new meteorological information provided by NNECO is consistent with the 1974-1975 Millstone data we have been using... recent licensing actions. Therefore, we can conclude that evaluations made using the 1974-1975 data are not changed.
2.10 Radiological Consequences of Postulated Accidents l
2.10.1 LOCA, Fuel Handling, and CEA Ejection Accidents We have reviewed the licensee's evaluation of the potential radiological I
consequences of the postulated LOCA, Fuel Handling Accident in the Spent Fuel Pool and CEA Ejection Accident in the SER dated May 10, 1974.
In addition, the postulated LOCA dose was reevaluated in our Reference 13 SE. All of these accidents were evaluated at a core thermal power level of 2700 MW. Using the new meterological data would result in doses less than those reported in the SER and Reference 13 SE for these accidents.
Therefore, we conclude that the potential consequences of these accidents are within the guidelines of 10 CFR Part 100 and are acceptable.
2.10.2 Main Steam Line Rupture, Steam Generator Tube Rupture and Waste Gas Decay Tank Failtre The May 10, 1974 SER states that the consequences of a steam line rupture, steam generator tube rupture and waste gas decay tank failure will be limited to small fractions of the dose guidelines of 10 CFR Part 100 by implementation of TS. These specifications are in effect at Millstone-2 and will not be affected by the power increase. Therefore, the conclusions reached in the 1974 SER are not changed by this action.
2.10.3 Fuel Handing Accident Inside Containment By letter dated January 18, 1977, we requested an evaluation of a postu-lated Fuel Handling Accident Inside the Containment (FHAIC). The licensee response concluded that the potential radiological consequences are within the guidelines of 10 CFR Part 100 (Ref.1). We have performed an inde-2131 162 A
c v.,
j
! i pendent analysis of the FHAIC. The assumptions used in our analysis are:
i Power Level (MWt) 2700 Fuel Exposure Time (yrs) 3
{
Equivalent number of assemblies damaged 1
Number of assemblies in core 217 Decay time before fuel movement (hrs) 72 Power peaking factor 1.65 l
Exclusion Area Boundary X/Q (Sec/cu m) 4.3 E-4 f
Other assumptions per Regulatory Guide 1.25 The results of our calculation show the potential consequences at the exclusion area boundary (EAB) are a dose of 94 rem to the thyroid and 0.4 rem to the whole body. The calculated potential consequences of the FHAIC are appropriately within the guidelines of 10 CFR Part 100 and are, i
therefore, acceptable.
" Appropriately within the guidelines of 10 CFR Part 100" has been defined as less than 100 rem to the thyroid. This l
1s based on the probability of this event relative to other events which are evaluated against 10 CFR Part 100 exposure guidelines. Whole body j
doses were also examined, but they are not controlling due to decay of the short lived radioisotopes prior to fuel handling. The potential consequences of this postulated accident at the Low Population Zone 4
i Boundary are less than those given for the EAB.
A recent study
- has indicated that dropping a spent fuel assembly into the core durinc refueling operations may potentially cause damage I
to more fuel pins than has 'been assumed for evaluating the FHAIC. This study has indicated that up to all of the fuel pins in two spent fuel assemblies, the one dropped and the one hit, may be damaged because of j
the embrittlement of fuel cladding material from radiation in the core.
The probability of the postulated FHAIC is small. Not only have there been several hundred reactor-years of plant operating experience with only a few accidents involving spent fuel being dropped into the core, l
but none of these accidents has resulted in measurable releases of j
activi ty. The potential damage to spent fuel estimated by the study was based on the assumption that a spent fuel assembly falls about 14 feet directly onto one other assembly in the core, an impact which j
results in the greatest energy available for crushing the fuel pins
{
- J. N. Singh, " Fuel Assembly Handling Accident Analysis," EG&G Idaho Technical Report RE-A-78-227, October 1978'.
2131 163
. in both assemblies. This type of impact is unlikely because the falling assembly would be subjected to drag forces in the water which should cause the assembly to skew out of a vertical fall path.
Based on the above, we have concluded that the likelihood of a spent fuel assembly falling into the core and damaging all the fuel pins in two assemblies is sufficiently small that refueling inside containment is not a safety concern which requires innediate remedial action.
We have, however, calculated the potential radiolooical consequences of a fuel assembly drop onto the reactor core with the rupture of all the fuel pins in two fuel assemblies. We have also assumed for this postulated accident that the source term for both spent fuel assemblies is that given in Regulatory Guide 1.25.
This is conservative because:
- 1) these two assemblies should not have the power peaking factor and clad gap activity recommended in Regulatory Guide 1.25 and 2) the pool decontamination factor for inorganic iodine should be greater than that recommended in Regulatory Guide 1.25. The calculated potential radiological consequences at the EAB for the complete rupture of fuel pins in two assemblies are twice the values given pre-viously (94 rem to thyroid and 6 rem whole body).
These conservatively calculated potential consequences are within the guidelines of 10 CFR Part 100; consequently, we have concluded that the potential consequences of the postulated FHAIC are acceptable.
2.10.4 Engineered Safety Features (ESF) Cc;nponent Leakage Outside Containment The licensee has not provided an analysis of the additional radiological consequences of leakage from the ECCS and Containment Spray System (CSS) during the recirculation phase follcwing a LOCA.
We have estimated the radiological consequences of ECCS and CSS component ledage based on a maximum leakage rate of 6 gallons / hour from each system.
The assumptions used in our analysis are:
e USNRC Standard Review Plan 15.6.5, Appendix B, " Radiological Consequences of a Design Basis Loss-of-Coolant Accident:
Leakage from Engic= red Safety Features Components Outside Containment."
e Core Power Level = 2700 MWt.
ESF Systems Recirculation Leakage is assumed to begin 30 minutes e
following the LOCA.
- e. ESF System Recirculation Leakage is assumed to be at a rate of 24 gallons / hour (twice maximum operational leakage),
e Recirculation Fluid volume = 500,000 gallons.
e Fraction of recirculation fluid flashing to steam = 0.l.
2151 164
. Overall charcoal filter efficiency for removal of iodines = 90%.
e X/Q for elevated release as given in Secti6n 2.9.
e Using these assumptions, we calculate a thyroid dose of 2.1 rem at the exclusion area boundary (0-2 hours) and 2.2 rem for the low population zone thyroid dose (0-30 days). The incremental doses of ECCS and CSS leakage, when added to the LOCA doses, are within the guidelines of 10 CFR Part 100 and are, therefore, acceptable.
The licensee has comitted to implement surveillance requirements on the maximum ECCS and CSS component leakage in the plant TS.
If the licensee proposes TS to limit the amount of leakage to values equal to or less than those assumed in our analysis, no further action will be required.
Due to the low probability of occurrence of a LOCA and the small consequences of this leakage, we will pursue this item separately and not require the specifications prior to the Cycle 3 operation.
2.10.5 Control Room Habitability Af ter LOCA The May 10, 1974, SER has evaluated the habitability of the control rcom following postulated accidents. The control room habitability systems were evaluated assuming a thermal power level of 2700 MW for postulated accidents. Therefore, the conclusion reached in the 1974 SER is rt changed by this action.
The 1974 SER identified a need for a chlorine detector to protect th: aperators in the event of a chlorine release. This detector has been installed, and its operability and surveillance are governed by the plant TS.
Based on the above, we conclude that the control room habitability systems are acceptable for operation at 2700 MWt.
2.10.6 Conclusion on Radiological Consequences of Postulated Accidents The potential ridiological consequences of design basis accidents have been previously evaluated at the proposed power level.
In addition, the potential consequences of a postulated FHAIC have been evaluated and found to be appropriately within the guidelines of 10 CFR Part 100 and are, therefore, acceptable.
2.11 Radioactive Waste Management The May 10, 1974 SER evaluates the liquid and gaseous waste treatment systems and the solid radioactive waste management system. The evaluation was performed assuming a thermal power level of 2700 MW.
Therefore, the conclusions reached in the SER are not changed 2131 165
. 2.12 Neutron Shield During the startup test program at Calvert Cliffs Unit No.1, higher than anticipated neutron levels resulting from neutron streaming in the reactor cavity were observed within the reactor containment building.
Design modifications were found to be necessary to reduce the dose rates below the design levels presented in the FSAR. Millstone-2 has a reactor design almost identical to that of the Calvert Cliffs Units and tests conducted during the Hillstone-2 startup test program confirmed that a similar condition existed at that facility. This condition resulted in limitations on containment access during reactor operations in order to minimize personnel radiatime exposure.
NNEC0 proposed L neutron shield design in Reference 10. After further review, the Ref erence 10 design was modified as discussed with the NRC staff on November 29, 1978.
NRC staff coments were included in the licensee's submittal of the final design (Ref.19).
2.12.1 Design Description The proposed reactor cavity neutron shield for Millstone-2 consists of two semi-circular compartmented water tanks.
When joined together the tanks form an annulus ring which covers the reactor cavity from near the vessel head to the refueling cavity floor.
Each of the tanks is subdivided into eight subcompartments and contain 16 inches of unborated water.
The tanks are constructed entirely of stainless steel.
The sides of the tanks and of each subcompartment are all one-half inch thick.
The tanks are sealed by means of one-eighth inch thick top and bottom cover plates which are welded to the radial and circumferential tank members.
An air gap provided between the surface of the water and the top cover plate allows water to expand inside the tanks and prevents over flow due to normal operational heatup.
Each tank is provided with redundant low pressure relief valves in order to minimize internal and external pressures ~
which arise as a result of thermal cycling.
Themal insulation and forced convective cooling are used to maintain the water in the tanks below 212 F.
The two semi-circular tank segments when placed in position within the reactor cavity are joined together by means of four hinge pins.
During the installation a "C Ring" which forms the lower I.D. of the shield structure engages the reactor flange.
Once the "C Ring" and hinge pins are installed, the shield structure is prevented from becoming a missile following a LOCA.
During the postulated LOCA event the bottom and top cover plates of the tank subc6mpartments rupture to provide the necessary cavity pressure relief.
The rupture is controlled by means of a series of grooves cut into the cover plates.
As the grooves tear, the panels collapse against the sides of the tank subcompartments to provide the necessary pressure r_elief area.
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. l 2.12.2 Structural and Mechanical Supporting arrangements and shield restraints, design, fabrication l
and installation procedures, structural analysis for all loads j
including seismic, LOCA pressure, dead load, thermal, lifting and I
handling,and miscellaneous loadings, load combir.atic75, stru:tural acceptance criteria, quality assurance were all reviewed in accordance with the criteria described in Standard Fevies Plan (SRP) 3.8.3, " Concrete.:.nd Steel Internal Structures of Steel or l
Concrete Containments."
The design, fabrication and erection of the structural steel is in compliance with the AISC, " Specification for Design, Fabrication and Erection of Structural Steel for Buildings",19E9.
Thermal, pressure, dead load, lifting and handling stresses were i
I analyzed utilizing a three dimensional model which covered a 90' segment of the tank with symmetry boundary conditions imposed at one l
end while the other end was leftunrestrained.
Sei snic. stresses on the shield due to the OBE and SSE loadings were analyzed.
Two analyses were performed in order to bound the seismic loadings.
In the first the entire tank assembly was allowed to slide through the maximum reactor flange to "C P.ing" diametral clearance, (2.0 inches) and a peak impact force was computed.
In the second the peak force was used to determine the maximum l
. acceleration of the water which further yielded an equivalent internal pressure force acting over various portions of the subcompa rtments.
A time history behavior of the burst panels under the action of the LOCA was performed to show when the i
panels fold against the sides of the tank and the effects of I
the LOCA on the rest of the shield structure. It was found that the panels so designed would lead to timely rupture and indeed j
provide the necessary cavity pressure relief during LOCA.
The si.resses derived from the above analyses were combined in I
accordance with the rules of S.R.P. 3.8.3 and found to be within the allowable stress limits.
l The criteria used in the analyses, design and construction of the neutron shield structure to account for anticipated loadings and postulated conditions that may be imposed upon the structure during its service lifetime are in conformance with established i
l criteria, codes, standards, and specifications acceptable to the NRC staff.
The use of these criteria provides reasonable assurance that the neutron shield structure will withstar.d the, specified design conditions without impairment of structural i
ir.earity or tne perfornance of required safety fun::ior.s.
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2.12.3 Asymetri c LOCA Loads l
HNECO tid not include an evaluation of the effects to the reactor We have determined that vessel due to asymmetric pressure loads.
the overall asymmetric LOCA loads an: lysis for Millstone-2 will not be adversely affected by the neutron shield and may be reviewed l
under the Task Action Plan A-2 which is currently scheduled to be completed by January 1,1980. We have,therefore, not included a review of this aspect.
2.12.4 Reactor Cavity Press _ure Transient _
I t
NNECO has rean alyzed the reactor cavity pressurization to verify the i
This reanalysis was based adequacy of the reactor cavity wall design.
We have reviewed on the double-ended guillotine break of the cold leg.
the reactor cavity nodalization and the input parameters used by the We concluded that i
licensee to perfom the pressurizatica analysis.
The licensee the analysis was done in a reasonably conservative manner.
utilized the computer code RELAP 4/ MOD 5 to calculate the pressuresa acceptable We have previously concluded that RELAP 4/5 ia transient.
analytical model for this kind of analysis.
i We have performed confimatory analyses which show that many of the peak pressures are approximately 20 percent higher than those calculated by the licensee and that for the break node, our calculcated pressure is about 30 percent higher than that calculated by the licensee.
Our analysis used the COMPARE subcompartment code and previously approved mass and energy release rates for a simila,' plant but wi t
higher power rating.this comparison is due to these higher energy release rates Although the surface-weighted peak pressure, expected at Millstone-2.
based on our results, would be larger than the licensee's value of 131 psi, the value appears to be well below the design pressure of 247 psig.
We, therefore, conclude that the licensee's analysis is acceptable.
j 2.12.5 Effects on post-LOCA Hydrogen Generation The tanks are constructed of stainless steel which could not generate a significant amount of hydrogen due to corrosion in a post-LOCA Therefore, we have concluded that the installation of environment.
this neutron shield will not cause an increase in the hydrogen generation j
rate following a LOCA.
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. 2.12.6 Containment Pressure The licensee analyzed the effect ;f the increased water inventory The a nunt of water stored in the shield on containment prassures.
ha about 2500 gallons at a tank was conservatively estimateo *t lne licensee concludes that temperature range of 160 F to 200 F.
the water has a negligible effect on calculate.1 containment peak pressure and containment backpressure for ECCS performance analysis.
We concur with the licensee's conclusion that the effect on the previous containment peak pressure and minimum pressure analyses is negligible.
2.12.7 Sump Clogging The neutron shield is designed to minimize the potential for clogging of the recirculation sumps which provide fluid for safety injection Sump clogging is precluded and containment spray following a LOCA.
by both the quantity of material which generates debris and by theTh location of the shield.
tank will not cause the flow blockage for safety injection and con-tainment spray recirculation following a LOCA.
2.12.8 Fire Hazard The materials used in the neutron shield are stainless steel, water, These materials are rock wool insulation and microtherm insulation.Therefore, we conclude that th all non-flammable.
does not present a fire hazard.
2.12.9 [:cupational Exposure Considerations We have reviewed the licensee's proposal for shielding the neutron streaming from tl<a reactor annular gap, as it affects occupational Based on the shield design, with at least 16 radiation exposure.
inches of water, NNEC0 estimates a dose reduction by a factor of 42.
The estimated neutron dose rate directly above the reactor cavity at the operating deck level would be reduced from 60 rems /hr to These dose approximately 1.4 rems /hr by the proposed shielding.
equivalent rates are based on neutron flux measurements taken at the reactor vessel flange level in the annulus, and using the DOT 3.5 and CASK codes developed at Oak Ridge National Laboratory.
The licensee estimates that 3 man-rems will be required for initial installation of the neutron shield, about 2 man-rems for rr3P56) and reinstallation of the shielding during an outage, and asatorad 0.7 man-rem resulting from storage of the shielding on v'r u fi e i L
.see of the steam generators during a refueling evtage.
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I has provided estimates of expected annual man-rem doses from all l
operations inside containment when the reactor is at power, and during refueling operations.
The total annual dose is estimated to be 120 man-rems /yr without the shield, and 20 man-rems /yr with the shield in place, in(uding expected additional activities I
in the containment made feasible by the reduced radiation levels.
i As a result of reviewing the shielding information provided, we conclude that the licensee's predicted man-rem doses with the shield in place are reasonable.
These doses represent a small part of the total annual occupational dose expected from all plant operations, and are acceptable.
Furthermore, taking into account i
dose rate measurements given in the Startup Test Report, we con-clude that the predicted dose rates, with the shield in place, both within the containment and in nearby occupied areas, will be consistent with the design dose rates in these areas, and l
are acceptable.
We conclude that the impact of this modification will be to l
effect a significant reduction in occupational radiation exposure
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to personrel entering the containment for inspection and minor maintenance while the reactor is at power and that their exposure will be as low as is reasonably achievable, and below the limits l
of 10 CFR Part 20.
NNEC0 has a; reed to provide an assessment of the neutron dose rate reduction and the actual man-rem exposure sr.ving experienced ciring Cycle 3 operation due to the instalfation of this neutron shield.
i 2.13 Containment Air Recirculation System Response Time In Reference 6, NNEC0 proposed to add a response time of less than or equal 31.0 seconds in TS Table 3.3-5 for the containment air recirculation (CAR) system. This time allows 20 seconds for the diesel generator to reach normal speed and voltage, 2.2 seconds i
for Sequencer Step 1 to close the fan breakers and the remaining 8.5 seconds for the CAR fan to reacr. normal speed. This response time will be an input parameter for the LOCA analysis. We j
find the 31.0 second CAR fan response time acceptable.
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2.14 Main Steam Line Safety Valves NNECO has proposed to modify the action statement of TS 3.7.1.1 to allow continued operation at reduced power level,with up to six inoperable main steam line code safety valves (Ref. 3).
Reference 18 cor-rected the Reference 3 calculated acceptable maximum power level to account for the increase in power level to 2700 MWt. The present Millstone-2 TS requires the plant to proceed to cold shutdown in the event one main steam line safety valve becomes inoperable.
Sixteen safety valves, eight on each main steam line, are provided to realize 108 percent of the steam generation rate for 2700 MWt power operation (11.8 X 106 lbs/hr).
This steam relief is necessary to maintain RCS heat removal under accident conditions when the main condenser is not available.
In other facilities with standard TS, we have utilized the following equation:
I SP = X - YV where:
SP = reduced reactor trip setpoint in percent of rated thermal power; Y = maximum number of inoperable safety valves per steam line; X = total relieving capacity of all safety valves per steam line; and Y = maximum relieving capacity of any one safety valve.
h Use of the above equation yields maximum power levels with less than all safety valves operable nearly equal to the value pro-posed by NNECO.
However, this equation and our proposed TS Table 3.7-1 require that the power level-high trip setpoint i
be reduced to insure an automatic reactor trip for this degraded i
mode of reactor operation.
The NNEC0 staff has agreed to this version of TS 3.7.1.1 presently used at other facilities.
We find that continued reactor operation with a few inoperable main steam line safety valves is acceptable when the power level-high trip setpoint is reduced in direct proportion to the steam relief capacity available.
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i 2.15 Environmental Qualification of Class IE Eouinment In response to I&E Bulletin 79-01, NNECO notified the NRC that 26 safety-related valves in the Millstone-2 containment use stem-j mounted 1imit switches (SMLS) where environmental cualification docu-inentation for the SMLS cannot be produced (Ref. 26).
In the required 14 day response, Reference 31, NNEC0 reports that replacement SMLS with the appropriate documentation are available and their intention
,is to replace all SMLS on the 26 valves during the current refueling outage.
l In Reference 44, NNEC0 states that four of the SMLS (SI-614, 624, j
634, and 644) will not be replaced during the Cycle 3 refueling outage. This is because of the unavailability of the replacement limit switches. The above valves are the block valves for the four i
ECCS safety injection (SI) tanks. The SMLS provide valve position i
annunciation (open/ closed) only; and control / indication functions l
are derived from other limit switches integral to the motor operator. TS 3.5.1 requires that the power to the SI tank valve operators be removed after these isolation valves are opened. The i
licensee has agreed that the SMLS on SI-614, 624, 634 and 644 will be j
replaced during the first unscheduled cold shutdown after September 15, 1979 (when replacement SMLS are estimated to be available). We find that this replacement schedule is acceptable since other valve position indication is available and TS 4.5.1 requires surveillance on a 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> schedule of SI tank operability.
i i
In a further response (Ref. 36), NNEC0 states that the auxiliary spray i
valve used to cool the pressurizer at low RCS pressure (2-CH-517) has l
a nonqualified solenoid.
The function of this valve has recently j
changed to be used as an alternate flow path to preclude postulated i
boron precipitation following a LOCA.
NNEC0 finds that this is a unique situation and commits to replace this solenoid on 2-CH-517 prior j
to the start of Cycle 3 operation.
i We find that the replacement of all environmentally unqualified SMLS and the solenoid on 2-CH-517 is a satisfactory resolution of our concerns expressed in I&E Bulletin 79-01.
We, therefore, consider this item resolved.
l 2.16 Containment Electrical Penetrations l
In Reference 39,NNEC0 submitted the results of Insulation Resistance (IR) tests performed on electrical penetration modules during the Cycle 3 refueling outage. The licensee's previous conunitment to perform additional testing of certain penetration conductors during this re-fueling outage is contained in our April 19, 1978 SE.
i During the 1978 Cycle 2 refueling outage, NNECO replaced 20 General Electric (GE) low voltage control penetration modules with new modules manufactured by Conax.
They also conducted a testina program of the balance of the penetration conductors to ide oify those conductors with IR less than 100 megohms serving safety related circuits or 50 megohms serving nonsafety related circuits. A number of conductors in the control element, drive mechanism (CEDM) penetration modules had IR muasurements less than 100 or 50 megohms.
As a consequence, it was decided that a survey of CEDM penetration modules would be performed during subsequent refueling outages.
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i The original 1979 Cycle 3 outage testing program was to retest 81 l
nonsafety circuit conductors which remained in service during Cycle 2 operation after having IR's between 50 and 100 megohms when tested during the Cycle 2 outage.
The test conductors consisted of 50 within the CEDM power penetrations and 31 within the CEA position indication penetration assemblies.
Results of this test revealed j
that of the 81 conductors comitted to be tested, 42 had IR readings that decreased, with 25 IR readings dropping below 50 megohms.
These test results identified a changing condition of conductor IR.
Con-I sequently, NNEC0 expanded the testing program to require a test of at least 10 percent of all penetration conductors of a like type to the i
low voltage control penetrations. This 10 percent sample would also concentrate on areas within modules where a low IR was identified in 1978, however, random samples of high IR conductors were also included.
The results of the IR testing noted that three penetration types (CEDM power, CEA position indication and instrumentation) showed changing IR conditions.
While 47 percent of the conductors in these types of modules were tested, only 3 percent exhibited a changing IR or an IR less than 50 megohms. Also, 70 percent of all low IR conductors identified were located in the CEA position modules which have the highest conductor density of 140 conductors per module.
Eighty-five percent of the nonsafety conductors requiring change were contained in the CEDM power and CEA position indication penetration assemblies.
No anomalies occurred during Cycle 2 operation in these systems.
In all instraces, nonsafety conductors identified with IR i
readings less than 50..egohms were placed with conductors having an i
IR greater than 50 megohms. We find this criteria acceptable.
There are six instrumentation penetrations containing seven modules per penetration 'or a total of 42 modules.
Each module contained 85 conductors of the 14 gage size. A total of 25 modules were tested i
during the extended 1979 test program.
Initial selection of modules to be tested was based on 1978 test results where an IR of less than 100 megohms had been identified.
If any conductors had IR's less j
than 50 megohms during the 1979 tests, a 100 percent check was i
performed on the entire penetration assembly. The selection of wires to be included in the test was determined by their close proximity to known low IR conductors which were abandoned in 1978.
Following the procedures above, it was determined that two modules were in i
l service containing conductors with IR's less than 50 megohms.
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All degraded circuits were transferred to acceptable conductors so as to estaDlish a minimum operating value of 50 megohms. In addition, j
. all safety related circuits were moved out of the two suspect modules and into modules which are.emonstrated to have all' conductors with IR's above 100 megohms.
We find this relocation of circuits acceptable.
The elevated contact resistance (within splices) which was identified l
during 1978 tests in the low voltage control modules as a sorce of heat, was not identified in the balance of penetrations tested both in 1978 and 1979.
It can be postulated that poor contact resistance and I
the resulting heating effects in the low voltage control penetrations originally accelerated the IR degradation which is now only beginning i
l to surface in the balance of like type penetrations.
The initial inspec-tion of a removed module in 1978 established tha' the IR degradation had 3
l occurred in the outer epoxy seal.
NNECO is involved in additional engineering activities to develop a long-term solution to eliminate the need for a test program that is repeated j
at each refueling outage.
One of the solutions being considered is the replacement of the original (GE) penetration modules with an improved i
type. We agree that a long-term solution to the yearly test program should be found.
The licensee has agreed to propose a permanent type repair of the electrical penetrations at least 90 days before the begin-l ning of the Cycle 4 refueling outage.
It should be noted that the degradation of penetration conductors has not affected the mechanical integrity of the penetration pressure seals.
The GE penetrations were constructed with four epoxy seals in each pene-tration module.
The outer seals provide the mechanical support for
{
the pigtail conductors.
A positive nitrogen gas pressure of 20 pounds was maintained on the penetration modules throughout the Cycle 2 operating period to detect leaking seals.
Experience during this period shows no abnormal leakage and no repair of the seals was required.
The monitoring of the modules for nitrogen leaks has demonstrated that the deterioration of the dielectric does not degrade the integrity of pressure boundary.
On the above basis we conclude that there is reasonable assurance that no failures will occur during Cycle 3 operation as was the case during Cycle 2 operation.
Ue, therefore, find the electrical penetrations of the containment acceptable for one cycle of operation.
The licensee has agreed to propose a permanent type repair of the electrical penetrations at least 90 days before the beginning of the Cycle 4 refueling outage.
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2.17 Steam Generator Surveillance Extensive modifications to the upper drilled support plates in both of the Millstone-2 steam generators were perfonned between November 1977 and February 1978. The purpose of the modifications was to relieve the constraint provided by the stiff parts of the tube support plate; thereby, reducing the susceptibility to continued tube denting and minimizing the progression of support plate deformation and cracking.
The modifi-cations perfonned were described in detail in a February 15,1978 sub-mittal from NNECO.
License Amendment No. 37 of the Millstone-2 operating license included the NRC evaluation of the modifications ard imposed supplementary inservice inspection requirements, at the end of Cycle 2 operation.
The required inservice inspection of the steam generators was performed during the current refueling outage.
The inspection included eddy current, gauging, and visual examinations.
No defective tubes (tubes with greater than 40% through wall degradation) or degraded tubes (tubes with 20%-40% through wall degradation) were discovered by eddy current testing of approximately 32% of the tubes in steam generator 1 and 26% of the tubes in steam generator 2.
Results of the gauging program revealed one tube in steam generator 1 and two tubes in steam generator 2 which would not pass a 0.540" probe..The visual inspection showed that the-general condition in the steam generators was similar to that existing after the repairs were performed during the last outage. The gap between the upper support plate and the shroud, established by the January 1978 rim cutting, appeared essentially unchanged.
However, several " loose parts" were present.
These loose parts were pieces of a tube support plate which had cracked away from the main part of the plate.
The largest piece was bounded roughly by a Scm by 3cm rectangle. This piece was found on the No.10 support plate and probably came from the No.11, upper drilled support plate.
Several cracked tube hole to flow hole ligamnts were observed around the rim of the upper support plate.
Corrective action taken by the licensee included plugging one tube in steam generator 1 and two tubes in steam generator 2 which did not allow passage of the 0.540" probe.
Two additional tubes were plugged,.
one whose hot leg plug had been inadvertently omitted during the 1978 plugging and one which had a suspected " weld leaker".
In addition, all the observed loose parts were removed from the generators.
The inspections conducted were in accordance with the program approved by the NRC in License Amendment No. 37 and provided a comprehensive evaluation of the condition of the steam generators.
The modifications made during late 1977 and early 1978 have had the expected result of decelerating the progession of tube denting.
The deceleration is evidenced by the results of the tube gauging program.
Although the progression 2131 175
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of tube denting has reduced substantially there is no evidence to indicate that corrosion of the carbon steel tube support plates has stopped.
Removal of the stiff portions of the support plates would be expected to result in a decrease in tube denting accompanied by an increase in support plate deforation and cracking because the modification makes it easier for the plate to deform than for the tubes to be dented.
This hypothesis is borne out by the discovery of continued cracking and breaking away of sections of the support plate. Continued use of the full flow condensate polishing system and careful control of feedwater purity will minimize the potential for continued support plate corrosion.' However, careful monitoring af the condition of the tube support plates must continue.
A loose parts evaluation based on testing and evaluation was discussed in the February 15, 1978 Northeast Nuclear Energy Company submittal referenced previously.
The evaluation bounded the worst case loose part which might be anticipated as a result of support plate cracking and concluded that loose parts of the size under consideration would not present any safety concern.
Results of the eddy current testing indicate that the generators are in good condition relative to other forms of degradation resulting in tube wall thinning and that the loose parts discovered in the steam generator did not cause any tube damage.
We have concluded that the Millstane-2 steam generators are acceptable for continued operation; however, close monitoring for denting and support plate degradation should continue. The licensee has agreed to provide a steam generator inspection program to evaluate the existence of tube denting and the integrity of support plates.
This inspection program should be submitted for our review at least 90 days before the shutdown for the Cycle 4 reload outage.
2.18 Piping and Support System The NRC sent I&E Bulletin No. 79-07 to all pressurized water reactors (PWR) licensees because the potential exists that certain safety related piping systems at some facilities may have been analyzed with a computer code which incorrectly sumed earthquake loads algebraically.
Reference 37 provides NNECO's response to Bulletin No. 79-07.
NNEC0 states that three engineering firms (Combustion Engineering, Teledyne Engineering Services and Bechtel Corporation) performed the seismic analyses on safety related piping for the original Millstone-2 design. The Reference 37 response indicates that the algebraic combination scheme of multiple earthquake excitation response was never used in the original design calculations.
Although the original design calculations used appropriate seismic analyses, NNEC0 states that six a:.
es performed "in-house" were.done using a computer program, ADL PIPE, which use'd algebraic summation techniques. The six analyses were performed for minor system modifications or for verification of piping integrity.
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In Reference 43, NNEC0 states that the six systems have been reanalyzed using SRSS techniques of combining responses due to earthquake motion. The code used was ADL PIPE, Revision 18, dated February 1977.
Although a listing of this code has as yet not been submitted, we have been assured by the originator of this code (A. D. Little, Inc.) that it
"
- O Wiconfomssto the requirements of Regulatory Guide 1.92 regarding inter-modal and intramodal ' methods of combination. We find this version of ADL PIPE provisionally acceptable and our benchmark verification analyses is completed on a generic b& sis.
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'5The six systems requiring, reanalysis were:
- e. Volume Control Tank Changing Bypass Lines e Nitrogen Addition System o Charging System e Diesel Generator Exhaust Piping e Reactor Coolant Pump Tap Root Valve Instrument e Safety Injection and Containment Spray Tes; Line The licensee states that all these piping systems, supports and attached equipment meet the applicable criteria of the plant FSAR.
In Reference 40, NHEC0 provides their response :o I&E Bulletin No. 79-04 on " Incorrect Weights For Saving Check Valve finufactured by Velan Engineering Corporation".
Their response provides assurance that no incorrectiVelan valve weights exist at Millstone-2.
I&E Bulletin No. 79-02 on " Pipe Support Base Plate Design Using Concrete Expansion Anchor Bolts", was not addressed during this re-evaluation since none of the piping supports exceed their allowable design limits.
The reanalysis technique employed was a lumped mass response spectra modal analysis.
The piping systems were modified as three dimensional lumped mass systems which included considerations for eccentric masses at valves and appropriate flexibility and stress intensification factors. The dynamic analysis procedures employed have been reviewed by the staff and are considered acceptable.
The results of the re-evaluation of these piping systems have been reviewed against the criteria of the applicable standards, ASME Section III for piping and support limits, and have been determined to be acceptable.
2131 177 s
3.0 Technical Specifications The TS changes proposed for this amendment are summarized in this section. All TS changes have been taken into account in generating the Cycle 3 setpoints and in the safety analysis.
Page 1-1 The definition of RATED THERMAL power would be changed from 256L MWt to 2700 MWt to reflect the proposed power increase.
4 Pt :s 2-2, 2-7, 2-8, 2-9 and B2-2 Revised Figures 2.1-1, 2.2-2, 2.2-3 and 2.2-4 are provided.
These figures, which relate to the TM/LP trip and the Local Power Density - High trip, are a result of the Cycle 3 setpoint 6nalysis. The setpoint analysis results are affected by the RATED THERMAL power, the core loading pattern, and the DNBR correlation used, all of which have changed in Cycle 3.
Page 2-4 Table 2.2-1 would be revised to place a "less than or equal" sign ahead of the minimum allowed Power Level-High setpoint.
This action is simply a correction for clarification.
Pages 2-5, 3/4 3-3, 3/4 3-6 and B2-8 Tables 2.2-1, 3.3-1, 3.3-2 and 4.3-1 would be revised to include the new underspeed - Reactor Coolant Pumps trip setpoint.
A basis for this trip will be specified.
Pages B2-1, B2-3, B2-5 and B2-6 These bases pages would be revised to reflect the use of the CE-1 DNBR corre'ation to replace the W-3 DNBR cor-relation.
The minimum allowed DNBR is reduced from 1.30 to 1.19.
Page B2-7 The time delay allowance (in units of pressure) for the TM/
LP trip would be increased from 30 psi to 45 psi to account for the revised CEA withdrawal transient computation per-formed in the Cycle 3 safety. analysis.
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_ 54 Page 3/4 1-26 The CEA drop time would be increased from 2.75 seconds to 3.1 seconds in TS 3.1.3.4.
This increase.in CEA drop time would result from the changed hydraulic characteristics of the four demonstration fuel assemblies with reduced CEA fl ow.
Page 3/4 1-30 Figure 3.1-2, the Power Dependent Insertion Limit (PDIL) would be revisad to allow less rod insertion than was allowed in Cycle 2.
The purpose of this is to maintain operating flexibility in the DNBR and LHR LC0's, the com-putation of which depends on the PDIL.
Pages 3/4 2-1, 3/4 2-2 and B3/4 2-1 The linear heat rate LC0 sealing equation would be modified to reflect the Cycle 3 analysis.
This analysis has absorbed the water hole peaking penalty factor and used a reduced measurement - calculational uncertainty factor.
Paaes 3/4 2-4 and 3/4 2-5 Figures 3.2-2 and 4.2-1, LC0 axial shape index and fuel gap augmentation factors, would be revised to reflect the power level increase and Cycle 3 analysis.
Pages 3/4 2-6, 3/4 2-8 and 3/4 2-9 The calctlated maximum value of Fxy and FT (total planar radial - ~d total integrated radial peaking factors) and Figure l'-3 would be modified as a result of Cycle 3 reanalysis.
Page 3/4 2-10 The applicability of TS 3.2.4 would be changed to not re-quire azimuthal power tilt determination until above 50 percent of rated power level.
This is consistent with other facilities with standard TS.
Pages 3/4 2-12 and B3/4 2-2 The TS limitation on cycle burnup would be removed in accordance with the fuel rod cladding collapse analysis pre-viously approved and reviewed in this SE.
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. Page 3/4 2-14 Table 3.2-1 would be revised to reflect the maximum allowed cold leg temperature of 549 F, consistent with the Cycle 3 safety analysis. The Cycle 2 value was 542.F.
Page 3/4 2-15 Figure 3.2-4, the Cycle 3 DNBR axial shape index LC0 curve, would be made more restrictive than the Cycle 2 curve.
This change is primarily due to a larger FT and higher power level.
Page 3/4 3-6 Table 3.3-2 would be revised to reduce the RTD response time from 10 to 8 seconds in accordance with the Cycle 3 analysis.
Pages 3/4 3-21 and 3/4 3-22 Table 3.3-5 would be modified to include an engineered safety feature response time for the charging pumps and for the containment air recirculation system.
Pages 3/4 4-5, 3/4 4-7a through 3/4 4-7h, B3/4 4-2 and B3/4 4-2a TS 3.4.5 and the related surveillance requirements and basis would be modified to remove the specifications for a steam generator supplementary inspection program.
Paaes 3/4 5-3, 3/4 5-4, 3/4 5-5 and 3/4 5-6 TS 3.5.2 and 4.5.2 would be modified to include LC0 and surveillance requirements for the use of the charging pump flow as an ECCS as required by the small break LOCA analysis.
Pages 3/4 7-1, 3/4 7-2, B3/4 7-1 and B3/4 7-2 TS 3.7.1.1 and its basis would be changed to allta continued operation with a reduced power level - high trip se tpoint when a few main steam line safety valves are inoperable.
Pages 3/4 7-21 and 3/4 7-22 A special exemption for inaccessible hydraulic snubbers, void after completion of Cycle 2', would be removed from TS
- 4. 7. 8.1.
2131 180
. Page 3/4 7-35 TS 4.7.9.1.2.c would be revised to remove :he words "during shutdown" from the 18 month surveillance requirement for the diesel fire pump.
This change is a correction since this surveillance is allowed to be performed by Action Statement a of TS 3.7.9.1.
Page 3/4 10-1 TS 3.10.1 will be made more understandable by the inclusion of a greater than or equal to indication before the 1720 ppm boric acid solution to be injected upon loss of shutdown margin.
Page 3/4 10-2 TS 3.10.2 would be revised to add TS 3.1.1.4 (moderator temper-ature coefficient values) to the list of specifications that may This would be suspended during the performance of Physics Tests.
correct the TS since the suspension is n-ild currently on page 3/41-5 (the page for TS 3.1.1.4).
4.0 Physics Testing The physics startup test program for Millstone-2 Cycle 3 was described in Reference 18.
The test program consists of hot functicnal tests, low power physics tests and power ascension tests.
The purpose of the startup test program is to provide assurance that the core conforms to the design analyzed.
The low power tests consist of CEA symmetry checks, critical boron concentration tests, CEA group worth tests and isothermal temperature coefficient tests. The power ascension tests consist of critical boron concentration tests, power distribution verification, isothermal temp-erature coefficient tests and power coefficient tests.
For each test, the acceptance criteria and actions to be taken if the acceptance criteria are not met were reviewed. More information has been provided on the acceptance criteria and the required actions in Reference 26.
This entire program, including the tests, the acceptance criteria and the actions has been reviewed and approved by the NRC staff.
5.0 Appraisal of Operating Effectiveness Based on the enclosed memorandum from Mr. G. R. Klingler of the NRC Office of Inspection and Enforcement, dated February 9, 1979, we find that no conditions adver se to safe operation of Millstone-2 at 2700 MWt have been identified.
2131 181 f.
=
6.0 Conclusion 6.1 Conclusion on Safety We have concluded, based on the considerations discussed above, that:
(1) for the items reviewed herein there is reasonable assurance that the heal and safety of the public will not be endangered by Cycle 3 operation of Millstone-2 at a core thermal power level of 2560 Mw, and (2) such activities will be conducted in compliance with the Commission's regulations and the issuance of this amendment will not be inimical to the common defense and Further, with the security or to the health and safety of the public.may require additional review exception of the items which the TMI-2 Accident, we conclude that operation up to a maximum of 2700 MWt would be acceptable.
5.2 Power Increase Review In accordance with previously established practices, a request for an increase in reactor power level is to be reviewed by the Reactor Operations Subcommittee of the Advisory Committee on Reactor Safeguards (ACRS). This subcommittee makes the determination in regard to a full committee review.
Such a subcommittee meeting was scheduled for May 9.1979, but as a result of the TMI-2 Accident, neither the NRC staff nor the ACRS The review could support the review of the power increase for Millstone-2.
meetings will be scheduled in the near future.
In the interi m, pending further staff and ACRS review, the power level of Millstone-2 will be restricted to 2560 MWt (94'.8% of the 2700 MWt power level approved by our analyses) by existing license condition 2.C.(1).
6.3 Environmental Consideration We have determined that the amendment for Cycle 3 operation at 2560 MWt does not authorize a change in effluent types or total amounts nor an increase in power level and will not result in any significant environ-mental impact.
Having made this determination, we have further concluded that this action involves an action which is insignificant from the standpoint of environmental impact and, pursuant to 10 CFR g51.5(d)(4),
that an Environmental Impact Statement or Negative Declaration and Environmental Impact Appraisal need not be prepared in connection with the issuance of the amendment for Cycle 3 operation at 2560 MWt.
We will. however, prepare an Environmental Impact Appraisal in connection w th the licensee's request to allow operation of Millstone-2 at increased i
po'..er l evels up to 2700 2t.
These documents will be issued concurrently with a f further Commission action concerning operation at this increased power "ecel.
Cate;.
2131 i32
TOPICAL REFERENCES CEN-89(N)-P, " Solution to Increased Water Hole Peaking in Operating a.
Reactors (Millstone-2)", D. C. Switzer to R. Reid, April 10, 1978.
b.
CENPD-98, "C0AST Code Description," April 1973.
c.
CENPD-107, "CESEC - Digital Simulation of a CE Nuclear Steam Supply System," April 1974.
d.
CENPD-132, " Calculative Methods for the CE Large Break LOCA Evaluation Model", August 1974, December 1974 (Supplement 1), and July 1975 (Supplement 2).
CENPD-133, "CE' FLASH-4A, A FORTRAN-IV Digital Computer Program for e.
Reactor Blowdown Analysis", April 1974, August 1974 (Supplement 1),
December 1974 (Supplement 2), and January 1977 (Supplement 3).
l f.
CENPD-134, "COMPERC-II, A Program for Emergency Refill-Reflood of the Core", April 1974 and December 1974 (Supplement 1).
CENPD-135-P, "STRIKIN II, A Cylindrical Geometry Fuel Rod Heat Transfer g.
Program," August 1974, February 1975 (Supplement 2-P), August 1976 (Supplement 4-P) and April 1977 (Supplement 5-P).
h.
CENPD-137, " Calculative Methods for the C-E Small Break LOCA Evaluation Model", August 1974 and January 1977 (Supplement 1).
- i. CENPD-138, " PARCH, A FORTRAN-IV Digital Program to Evaluate Pool Boiling, Axial Rod and Coolant Heatup", August 1974, February 1975 (Supplement 1) and January 1977 (Supplement 2).
- j. CENPD.a!, "CE Fuel Evaluation Model", July 1974.
k.
CENPD-145, " INCA: Method of Analyzing In-Core Detector Data in Power Reactors", April 1975.
1.
CENPD-153, " Evaluation of Uncertainty in the Nuclear Form Factor Measured by Self-Powered Fixed In-Core Detector Systems",
August 1974.
CENPD-161-P, " TORC Code - A Computer Code for Determining the Thermal m.
l Margin of a Reactor Core," July 1975.
n.
CENPD-162-P-A, "CE Critical Heat Flux", September 1976.
CENPD-183, "CE Methodr. for loss of Flow Analysis", July 1975.
o.
1
?
2131 W I
i 1
1 CENPD-187, "CEPAN Method of Analyzing Creep Collapse of Oval Cladding",
i p.
June 1975.
CENPD-190A, "CEA Ejection, C-E Method for Control Element Assembly 9
Ejection", July 1976.
CENPD-199-P, "C-E Setpoint Methodology", April 1976.
r.
WAPD-TM-479," CHIC-KIN - A Fortran Program for Intermediate and Fast1965.
s.
Transients in a Water Moderator Reactor",.l. A. Redfield,.lanuary WAPD-TM-678,"PDQ-7 Reference Manual", W. R. Cadwell, January 1978.
I t.
l WAPD-TM-743, "TWIGL - A Program to Solve The Two-Dimensional, Two Group.
Space-Time Neutron Diffusion Equations with Temperature Feedback",
u.
J. B. Yasinsky, M. Natelson, and L. A. Haaemen, February 196R.
j
" Interim Safety Evaluation Report on Effects of Fuel Rod Bowing on Thermal Margin Calculations for Light Water Reactors", Pages 21 and 26 I
v.
(NRC Report).
CENPD-136, "High Temperature Properties of Zircaloy and UO2 for Use w.
in LOCA Evaluation Models," July 1974.
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2131 184 l
~
USE OF TOPICAL REPORTS Evaluation Topicals Used CYCLE 3 CORE 2
4 5
l a, k,5,7,m,n,p,r Core Parameters Clad Collapse p
V Fuel Rod Bowing 3
- I '4
, c,5, m, n, v a,5 2
ANTICIPATED OPERATIONAL OCCURRENCES
- I'4 a,5, c ', m, n, r CEA Withdrawal
- I RCS Depressurization a ', c ', m, n, r c,5
- I 3
,r Boron Dilution Loss of Load FSAR*, c, m, n Loss of Feedwater Flow FSAR*, m, n l
Loss of Coolant Flow b, c ', g, m, n, o, r
- 1 CEA Drop c,m,n,r -
Transient Resulting from c,5
- )
3 Malfunction of One Steam Generator
, m, n, r ACCIDENTS CEA Ejection q,s,t,u FSAR*, c,5, m, n 3
Steam Line Rupture FSAR*, c,5 3
, m, n Steam Generator Tube Rupture 3
l FSAR*, c,5,g,m,n,o RCP Seized Rotor LOSS OF COOLANT ACCIDENT I
Rupture Strain Model w
Large Break FSAR*, d, e, f, g, i, j Small Break FSAR*, e, f, 9, h, i
- Indicates the location of the most complete description of the event and its associated analytical methodology.
2131 185 Table Key:
Letters - See Topical References Numbers - Refers to Notes on Next Page
- 61 Notes:
All topical reports have been approved by the NRC staff except as noted with the following superscripts, i
Staff review has progressed to the point where this 1.
methodology is considered to be acceptable for reload analysis.
Under staff review; TM/LP methodology is acceptable.
2.
Under staff review; staff comparison of CESEC results 3.
with RELAP-3 computations shows good agreement; staff l
considers accept'able for reload analysis.
l Under staff review; INCA methodology is acceptable, l
4.
Uncertainty analysis has been superseded by subsequent S.
CE analysis (Ref. 22 ).
l 1
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2131 186 e9 N
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- m - e - -
.r~ -
I i
LETTER REFERENCES
?
NNEC0 provides evaluation of postulated fuel handling accident inside I
1.
containment, D. Switzer to G. Lear, March 21, 1977.
l 2.
NRC issues Amendment No. 26 allowing an increase in discharge delta temperature, G. Lear to D. Switzer, May 18, 1977.
l 3.
NNECO applies for inoperable main steam line safety valves TS, D. Switzer to G. Lear, December 16, 1977.
NRC issues Amendment No. 38 for the Cycle 2 reload, R. Reid to D. Switzer, 4.
April 19,1978.
5.
NNEC0 provides information on waterhole peaking and indicates future power level increase, W. Counsil to R. Reid, July 28, 1978.
NNEC0 provides information on diesel generator start times, W. Counsil l
6.
i to R. Reid, October 11, 1978.
NRC issues Amendment No. 18 for the Cycle 2 reload of Calvert I
7.
l Cliffs Unit No. 2, R. Re % to A. Lundvall, October 21, 1978.
t NNECO notifies of intention to increase power for Cycle 3 8.
i operation, W. Counsil to R. Reid, November 1,1978.
I NNEC0 presents design of RCP speed sensing system, H. Counsil 9.
to R. Reid, November 8,1978.
i NNECO presents design of neutron shielding, W. Counsil to R. Reid, 10.
November 13, 1978.
f
- 11. NRC summary of November 21, 1978 meeting on power increase program, E. Conner to file, January 3,1979.
- 12. NRL summary of November 29, 1978 meeting on neutron shield design, E. Conner to file, December 21, 1978.
f NRC issues Amendment No. 45 for miscellaneous TS changes, R. Reid f
13.
j to W. Counsil, December 8, 1978.
I
- 14. NNECO applies for Cycle 3 reload / power increase, W. Counsil to R. Reid, December 15, 1978.
NNECO presents CEA guide tube inspection program, W. Counsil to 15.
l R. Reid, December 18, 1978.
NNECO provides TS revisions for Cycle 3 at 2700 MWt, H. Counsil t
16.
to R. Reid, December 28, 1978.
NNECO presents charging pump qualification, W. Counsil"to R. Reid, 17.
February 6,1979.
18 NNEC0 applies for power increase, W, Counsil to R. Reid, February 12, 1979.
NNEC0 provides neutron shielding design qualification, W. Counsi1 19, to R. Reid, February 23, 1979.
2131 187 i
j
20.
NNECO provides analysis of three incidents W. Counsil to R. Reid, February 7,1979.
21.
NNEC0 applies for RCP speed sensing system TS, W. Counsil to R. Reid, March 2, 1979.
22.
CE presents data justifying measurement uncertainties of 6 percent in Fr and 7 percent in Fq, A. Scherer to P. Check, March 7,1979.
23.
NRC requests additional information on Cycle 3 reload /powe.r increase, R. Reid to W. Counsil, March 14, 1979.
24.
NNECO provides non-QA version of snall break LOCA analysis, W. Counsil to R. Reid, March 22, 1979.
25.
NNEC0 provides trip setpoint for RCP speed sensing system, W. Counsil to R. Reid, March 23, 1979.
26.
NNECO responds to I&E Bulletin 79-01 on environmental qualifications of Class lE equipment, W. Couns1; to R. Reid, March 23, 1979.
27.
NNEC0 provides additional information as requested on March 14, 1979, W. Counsil to R. Reid, tbrch 27, 1979.
28.
NNECO provides results of new RTD test method, W. Counsil to R. Reid, March 28,1979.
29.
NNEC0 provides sleeved CEA guide tube inspection program, W. Counsil to R. Reid, March 29, 1979.
30.
NNEC0 provides non-QA version of large break LOCA analysis, W. Counsil to R. Reid, March 30, 1979.
31.
NNEC0 provides environmental qualification of Class IE equipment, W. Counsil to R. Reid, April 6,1979.
32.
NNECO certifies that small break and large break LCCA analyses (Refs. 23 & 29) have QA verification, W. Counsil to R. Reid, April 9, 1979.
33.
NNEC0 provides seismic qualification of RCP speed, sensing system, W. Counsil to R. Reid, April 12, 1979.
34.
NNEC0 reports the steam generator inspection results, W. Counsil to R. Reid, April 13, 1979.
35.
NNEC0 reports the CEA guide tube inspection results, W. Counsil to R. Reid, April 17, 1979.
35.
NNEC0 provides further response in regards to environmental qualifications of Class IE equipment, W. Counsil to R. Reid, April 18, 1979.
6 2131 188 M
e s
a
, l l
t 36.
NNEC0 provides further response in regards to environmental qualifications of Class lE equipment, W. Counsil to R. Reid, 4
April 18,1979.
37.
NNEC0 provides response to I&E Bulletin No. 79-07, W. Counsil to B. Grier, April 24, 1979.
o 38.
NNECO reports further CEA guide tube inspection results, W. Counsil to R. Reid, April 26, 1979.
- 39. NNEC0 provides results of electrical penetration testing, W. Counsil to R. Reid, April 26, 1979.
40.
NNEC0 provides response to I&E Bulletin No. 79-04, W. Counsil to B. Grier, April 30, 1979.
41.
NNECO provides additional information concerning Cycle 3 safety analysis, W. Counsil to R. Reid, May 3,1979.
I i
42.
NNEC0 reports more CEA guide tube inspection results, W. Counsil to R. Reid, May 7,1979.
43.
NNECO provides further response to I&E Bulletin No. 79-07, W. Counsil to B. Grier, May 7, 1979.
44.
NNEC0 provides additional response in environmental qualification of Class lE equipment, W. Counsil to R. Reid, May 8,1979.
l 45.
NNEC0 provides additional information in regards to I&E Bulletin No. 79-02 and 79-07, W. Counsil to R. Reid, May 11, 1979.
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2131 189
.