Forwards Info Supporting Applicability of WCAP-11145, Westinghouse Small Break LOCA ECCS Evaluation Model Generic Study W/Notrump Code, to Facility,Per Bj Youngblood 861222 Request.Closeout of NUREG-0737 Item II.K.3.31 Requested| ML17334B086 |
| Person / Time |
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| Site: |
Cook  |
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| Issue date: |
04/23/1987 |
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| From: |
Alexich M
INDIANA MICHIGAN POWER CO. (FORMERLY INDIANA & MICHIG |
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| To: |
Harold Denton
NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM) |
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| References |
|---|
| RTR-NUREG-0737, RTR-NUREG-737, TASK-2.K.3.30, TASK-2.K.3.31, TASK-TM AEP:NRC:0916Z, AEP:NRC:916Z, GL-83-25, GL-83-35, NUDOCS 8704300061 |
| Download: ML17334B086 (145) |
|
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Text
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ACCESSION NBR: 8704300061 DOC. DATE: 87/0%/23 NOTARIZED:
NO DOCKET FAC IL: 50-316 Donald C.
Cook Nuc lear Pouer Plant'nit 2>
Ind iana Pi 05000316 AUTH. NAME
,AUTHOR AFFILIATION ALEXICHiM. P.
Indiana 5 Michigan Electric Co.
RECIP. NAME RECIPIENT AFFILIATION DENTONi H. R.
Document Control Branch (Document Control Desk) 4
SUBJECT:
Forwards info supporting appli,cab ilitg of lr1CAP-,11145>
"Mestiqghouse. Small Break,LON EGGS:Evaluation/Model Generic:
Studg -N/NOTRUMP'Code> '&to faci itgi per BJ Youngblood 861222 request. Closeout of NUREQ-0737 Item II. K. 3. 31 requested.
$&.p DISTRIBUTION CODE:
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OR Submittal:
TMI Action Plan Rgmt NUREG-0737 8c, NUREQ-0660 NOTES:
RECIPIENT ID CODE/NAME PD3-3 L*
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Attachment 1 to AEP:NRC:0916Z Page 2
AEP'NRC'0860B Now Unit 1 Westinghouse Exxon (now ANFC) 80 113 159 34 Unit 2 Westinghouse Exxon (now ANFC) 29 164 192 Use of Unit 1 small break LOCA analyses to support requested Unit 2 licensing actions was found acceptable in the Safety Evaluation Report (SER) prepared by the NRC staff for Amendment 64 to the Unit 2 license.
Applicable pages from this SER are included here as Exhibit C.
This acceptability for Unit 2 was further cited. in the SER for Amendment 84 to the Unit 1 license, included here as Exhibit D.
(Unit 1 Amendment 84 granted safety-injection mini.flow T/S changes identical to those granted to Unit 2 in Amendment 64).
More recently, ANFC performed a complete review of accidents cited in the NUREG-0800 Standard Review Plan Chapter 15.
This appeared in report XN-NF-85-28(p), Supplement 1,
"D.C. Cook Unit 2, Cycle 6 Safety Analysis Report:
Disposition of Standard Review Plan Chapter 15 Events."
This report was transmitted to you directly by ANFC in their letter RAC:069:85, dated October 15,
- 1985, and was referenced in our letter AEP:NRC:0916G, dated October 18, 1985.
Pertinent pages from this report are pages 137 through 143.
1
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Attachment 1 to AEP:NRC:0916Z Page 3
In this report, the applicability of the above cited Unit 1 analysis to Unit 2 was proposed once again.
In response, the SER for Amendment 82 to the Unit 2 license made no adverse findings in regard to this proposal.
That same SER reviewed the ANFC large break analysis in detail, and concluded that the 10 CFR 50.46 criteria are satisfied with no qualification.
Applicable pages from that SER appear as Exhibit E to this attachment.
Our review of the Unit 2 small break LOCA licensing basis has determined that the limiting small break LOCA analysis for D.
C.
Cook Unit 2 fueled with Exxon 17 x 17 fuel is a small break LOCA analysis for Unit 1 fueled with Westinghouse OFA 15 x 15 fuel.
For plants with Westinghouse fuel, the generic NOTRUMP analysis is accepted on the basis that the previous small break analysis method for the same conditions has been shown to be conservative by comparison.
The current analysis for D.
C.
Cook Unit 1 fueled with Westinghouse 15 x 15 OFA fuel was approved on this basis in a letter from Mr. B. J. Youngblood of the NRC staff to Mr. John Dolan, Vice Chairman of AEP, dated December 22, 1986 (as discussed in the cover letter for this submittal).
It follows that the generic NOTRUMP analysis has demonstrated that the existing analysis for Cook Unit 2 is conservative because the approved D.
C.
Cook Unit 1 analysis has been found to be applicable by the NRC staff to D.C.
Cook Unit 2 fueled with Exxon 17 x 17 fuel.
Therefore, we conclude that the NOTRUMP generic analysis should be applicable to Cook Unit 2.
Exhibit A of Attachment 1
to AEP:NRC:0916Z March 15, 1984 Letter from M. P. Alexich (I&NECo) to H. R. Denton (NRC) Regarding Unit 2 Cycle 5
Reload Technical Specifications
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IL.
7 B.
Safet In'ection Miniflow Line Modification Back round The Indiana 5 Michigan Electric Company
{IHEC0) submitted a request to modify the piping geometry of the miniflow line 'or the O.C.
Cook Unit Ho.
2 safety injection pumps by letters dated March 1 and 15, 1984.
- his modification has been previously performed for O.C.
Cook Unit Ho.
1.
As pre'sently confiaured, the miniflow line for Unit 2 is comprised of both 1.5 inch and 0.75 inch diameter piping.
The licensee has reauested that i be allowed to replace the 0.75 inch diameter piping with 1.5 inch piping, thereby making the entire piping in system of one diameter.
The purpose of this modification I
is based on economic and maintenance considerations.
By-maintain-ing both Units 1 and Z as similar as possible, the licensee is able, in many cases,
.o apply one analysis to both units.
increasing the miniflow line piping diameter doubles i s flow rate from 30 aom to (0 apm.
The increased floe is beneficial to the S[ pump when operating in the shut-off configuration in that it reduces he temperature rise through the pump.
This provides an added benefit of increased pump reliabili.y by allowina smoother l
operation at reduced temperatures.
increasing the miniflow coolant rate has a negative influence on ECCS performance in that it reduces the injected flow to the reac or coolant system.
At runout conditions-,
the ECCS injection rate is decreased from 63.0 ibm/sec to 61.6 ibm/sec.
At the other V
- extreme, the ECC injection rate at 1314.7 psia is reduced from 19.0 1bm/sec to 16.1 1bm/sec.
Since only the SI is influenced by the proposed hardware modfficatipn, the impact on large break LOCAs is
~t insignificant (total ECCS flow, not including accumulator injec-
- tion, is reduced from 463.0 ibm/sec to 461.6 ibm/sec).
This ~ould have negligible impact on the calculated peak clad temperature for the large breaks.
For the limi ing small break LOCA, however, IIlECO has determined that the peak clad empera ure would increase by about 87'F. 7'~
analysis was conservatively calculated for Unit 1, and was 'submit"='..
ted by INECO as applicable to Unit 2.
To demonstrate that the temperature increase
-,or Unit 1 was appli-cable to Un',
Il'.ECO had the reactor v ndor (Westinghouse) confirm.ha the ECCS pump characteris-'ics ror both Unit 1 and Onit 2 are identical.
Having anticipated the desirability to modify the geometry oi the miniilow line or Unit 2 as well, the limiting small break LOCA -.or Unit 1 was analyzed at, the Unit 2 power rating (3411 HWt versus 3250 ViMt).
In addition, the linear peak heat generation rate was analyzed at 16.67 kw/. for Unit 1 (Unit 2 is rated at 12.88 kw/it)'ince the linear heat generation rate for Unit 1 is significantly greater than that for Unit 2,
.he calculat-ed heat up rate would be conservative when applied to Unit 2.
The applicability oi the Unit 1 calculation to Unit 2 was also based on comparison of the volumetric fuel heat generation rate for the total core.
The volumetric heat generation rate for Unit 1 was
e E~p, 4
3 r
calculat
.-at;.9887 kw/ft of fuel and for Unit 2 at 9835 kw/ft of fueT.
Thcj otal volume of coolant in the core was also calculated to be nearly identical (614.8 ft and 613.0 ft for Units 1 and 2, 3
3 respectively).
With respect to the r maining primary system coolant vo'tume, both plants are identical.
The reduction of ECC injection by the SI pump resulted in an additional 6 inches of calculated core uncovery (5.5 ft ursus 5.0 ft).
This corresponded to a
10 second delay in coolant recove~ of the core (838 versus 848 seconds).
Tbo."nnsequential increase in peak clad temperature was 87'F.. With the present calculated sttiall" break peak cle ad temperature of 1668'F, the Unit' core response'for the limiting smatT: break LOCA is expected to be less than-1750'F.
This is well below the 2200'F 1icensing limit.
CONCLUSION OF THE MINIFLOM LINE REYIEM Me have review'ed the submi.tal by ttie Indiana 5, Nicii'. -;n'-Ele~iric Company to increase the pipe diameter of the miniflow line to the injection pumps.
The acceptability of 'he miniflow line mndi ficatin~. is based on the Unit 1
LOCA analysis and it's applicability to Unit 2.
"e find the onalysis and applicability acceptable, and therefore find acceptable, the requested modification of increasing the cross sectional diameter << 'he miniflow line from 0.75 inch to 1.5 inch.
We requested,
- however,
'.hat the Sl pump flow characteristic be corfi rmed to be consistent with the analys'.s assumptions prior to full power operation.
The lie.rsee has agreed tn perform this test prior to startup.
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+ygy4 UNlTEDSTATES NUCLEAR REGULATORY COMMISSlON WASHIMOTOM,O. C. %555 SAFETY EVALUATION SY THE OFFICE OF'NUCLEAR REACTOR RE6ULATION RELATED TO AMENDMENT NO.
82 TO FACILITY OPERATING LICENSE NO. DPR-74
,INDIANA'ND MICHIGAN ELECTRIC COMPANY DONALD C.
COOK NUCLEAR PLANT UNIT NO.. 2 DOCKET NO. 50-316 Introductf on Sy letters dated March 14 and March 27, 1986, the Indiana and Michigan Electric Company (the licensee),
proposed changes to the Donald C.
Cook Nuclear Plant, Unit No. 2.
These changes are grouped and evaluated below by.those changes re-lated to the Unit 2 cycle 6 reload and Technical Specification changes.
C cle 6 Reload The reactor core for D. C.
Cook Cycle 6 will contain 191 Exxon fuel assemblies and one Westinghouse fuel assembly each having a
17 x 17 fuel rod array.
Eighty eight of the Exxon fuel assemblies are new.
Cycle 6 burnup has been projected to be 17,790 Mwd/MTU at a core powe~ of 3411 MWT.
The design'haracteristics of the Exxon fuel assemblies were reviewed and approved by the staff for the cycle 5 core.
The Exxon fuel fn'the cycle 6 core will be of the same design as that of cycle 5.
As additional conffrmatfon of the fntegrfty of Exxon fuel remote visual examinations of irradiated fuel were performed following cycle 4.
No evidence of wear, fretting or other physical damage was noted for this fuel.
The exposure for the examined fuel ranged from 16.440 MWD/MTU to 17,630 MWD/MTU.
In anticipation of steam generator tube plugging the cycle 6 safety analyses were performed with an average steam generator tube plugging of 10%.
The effect of asyItletrfc tube'plugging was considered fn the analyses.
Fuel Thermal-Mechanfcal Desi n
Cycle 5 contafned a mixture of Exxon and Westfnghouse fuel elements.
Staff conclusions regarding the thermal-mechanical design of the cycle 5 core remain applicable to cycle 6.
In particular cladding strain, external corrosion (oxfdatfon), fuel rod fnternal pressure, and fuel rod pellet temperature were analyzed using the RODEX 2 code which has been approved by the NRC staff.
The analytical results satisfied the acceptance criteria.
Collapse of fuel cladding into a pre-existing axial gap produced by the differential pressure between the reactor coolant pressure and the internal fuel rod pressure was investigated.
It was determined that the cladding would not collapse.
cycle 'resul,t. was determfned to be bounding.
The pressurizer. pressure at.
beginning of cycle was adjusted to the maximum reactor system pressure location
.at the pump discharge and determined to be 2747.6.,psfa
.which fs less than 110%
of design.
Regulatory Guide 1.77 recomnends that the maxfaam pressure be less
- '.than that which would cause stresses to'xceed "Service Lfmft C" of, the ASHE Code.
'Service Lfmft C" corresponds to approximately 120% of design pressure.
The licensee's pressure calculation fs therefore acceptable.
The number of fuel rods experiencing DNBR less than the 1.17 design limit for EXXON fuel was calculated to determine the offsfte dose consequences from the event.
This was accomplished usfng the XCOBRA-IIIC code to determine the local power level at which the mfnfmgn DNBR would be 1.17.
A flow penalty was in-cluded to account for non-syametrfc effects.on core flow.from rod ejection.
Using the calculated radial power distribution f'r the XTRAN calculations the fraction of the core with local power levels greater than that which would produce a
DNBR of less 1.17 was determined.
From this calculation 10.7% of the fuel was predicted to penetrate the DNBR limit and to fafl.
The offsfte dose consequences from the event were calculated to be well within the exposure.
guidelines of 10CFR100 and are therefore acceptable.
Loss of Coolant Accidents Large break LOCA/ECCS analysis were performed fn 1982 (1, 2) to support operation of the D.C.
Cook Unit 2 reactor at 3425 HMT with ENC fuel.
The lfmftfng break was identified as the 1.0 double ended cold leg guillotine (DECLG) break as developed fn reference 1.
The results of calculations with one and two LPS pumps operatfng were presented fn reference 2 which indicated that a higher PCT occurred with two LPS pumps operation.
Reference 3 documents the results of LOCA/ECCS analysis performed in support of cycle 6 and future cycles with all ENC fuel at a thermal power rating of 3425 HMt, with up to 10K of the steam generator tubes plugged.
Calculations were performed for the previously identified 1.0 DECLG break, with full ECCS flow.
Three exposures using a center peaked axial power shape were studied to determine exposure dependence.
The exposures range from 2 HMD/kg to 47 HMD/kg peak rod average burnup.
The axial dependence of the peaking factor limit fs denoted K(Z) and fs defined as (K(Z) ~ Fg(Z)/HAX Fg(z) where Fg(Z) fs the maximum peaking factor allowed at any elevation Z.
The topmost segment of the K(Z) curve fs limited by the small break LOCA linear heat generation rate (LHGR) limits presented in the technical specifications.
Confirmation of the axial dependence fs based on three power distributions:
a center peak chopped cosine power dfstrfbutfon and two conservative top skewed power shapes as presented in Figure 3.P'.
The power distributions are analyzed at the limiting exposure, 2 HMD/kg, where the peak stored energy occurs.
A summary of these results and the exposure study fs presented fn Table 3.5*.
The normalized K(Z) curve verses core axial height fs presented in Ffgure 2.1*.
The calculations were performed using the EXEH/PMR LOCA/ECCS models, including Reference 3, Exxon Report XN-NF-85-68(P) Revision 1.
I 4
IF I'
4
~
~
'I 4 fuel properties calculated at the. start of the LOCA transient with the NRC,,
generically approved RODEX2 code.(4).
The. quench. tfme, quench velocfty and CRF correlations fn REFLEX and the heat transfer correlation fn TOODEE2 are based on ENC's 17x17 Fuel Cooling Test Facility (FCTF) data (5, 6; 7).
Reference 3
.reflects the revisions fn the correlations based on the FCTF data which are documented fn References 6 and '7 and to reflect a 1. 1 mltfplfer on the peak power parameter used 'fn the FCTF correlations fn the REFLEX code.
This documents the NRC acceptance of the 1.1 aaltfplfer as applied and agreed to with EXXON Nuclear Company, Inc.
The NRC finds that the analysis results and methods sumnarfzed above and as presented fn reference 3 supports operation of -the D.C. Cook Unit 2 reactor for cycle 6, and future. cycles with ENC fuel, at a total power peaking factor limit (Fg) of 2.10.
Consfderfng the results of the analysis as summarized fn Table 3.5*, peak+lad temperature fs less than 2200'F. local oxidation fs less than 17% and core wide metal-water reaction fs less than 1.t5.
Therefore we find that the criteria of 10CFR50.46 have been satisfied.
REFERENCES Loss of Coolant Accidents 2.
XN-XF-82-35. 'Donald C.
Cook Unit 2 LOCA/ECCS Analysis Using EXEM/PWR Large Break Results,'xxon Nuclear Company, Inc., Richland, WA
- 99352, Aprf1 1982.
'I XN-NF-82-35, Supplement 1, "Donald C.
Cook Unit 2 Cycle 4 Lfmftfng Break LOCA/ECCS Analysfs Usfng EXEM/PWR," Exxon Nuclear Company, Inc., Rfchland, WA 99352, November 1982.
3.
'Donald C.
Cook Vnft 2 Limiting Break LOCA/ECCS Analysis, Revision 1, 10%
Steam Generator Tube Plugging, and K(z) Curve," XN-NF-85-68(P) Revision 1, Exxon Nuclear Company, Inc., Richland, WA 99352, April, 1986.
4.
XN-NF-81-58(P)(A), Rev. 2, and Rev.
2 Supplements 1 and 2, "RODEX2: Fuel Rod Thermal-Mechanical
Response
Evaluation Model," Exxon Nuclear Company.
Inc., Rfchaldn, WA 99352, February 1983.
5.
XN-NF-85-16(P), Volume 2, PWR 17x17 Fuel Cooling Test. Program Reflood quench Carryover, and Heat Transfer Correlations,"
Exxon Nuclear
- Company, Inc., Richland, WA 99352, May 1985.
6.
"PWR 17x17 Fuel Cooling Test Program Reflood quench, Carryover, and Heat Transfer Correlatfons,"
XN-NF-86-16(P), Revision 1, and all supplements, Exxon Nuclear Company, Inc., Rfchland, WA 99352, January 1986.
7.
"PWR 17xl7 Fuel Cooling Test Program Sensitivity Studies,'N-NF-85-16(P),
Volume 1, and all supplements, Exxon Nuclear Company, Inc., Richland, WA 93352, January 1986.
Reference 3, Exxon Report XN-NF-85-68(P) Revision 1.
ATTACHMENT 2 to AEP:NRC'0916Z Detailed Evaluation of the Impact of Nuclear Fuel Assembly Design on NOTRUMP Small Break LOCA Analysis Results
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Attachment 2 to AEP:NRC:0916Z I
Page 1
Attachment 2
An evaluation of the impact of nuclear fuel assembly design on NOTRUMP small break LOCA analysis results can be established from work recently performed by Westinghouse at the request of American Electric Power.
This work was done in support of an effort to change our licensing basis for operation of the safety injection system.
Specifically, the work evaluated the acceptability of operating with the safety injection system cross-tie valves closed, such that each of the two safety injection pumps would only be capable of delivering flow to two Reactor Coolant System (RCS) cold legs.
(The previous small break analyses required the cross-tie valves to be open, such that each pump would be capable of providing flow to all four RCS cold legs.)
An evaluation to support operation of D.C.
Cook Unit 2 with the safety injection cross-tie valves closed was submitted in our letter AEP:NRC:1024, dated March 23,
- 1987, which we have included as Exhibit B to this attachment.
Included as Exhibit C to this attachment are excerpts from a letter from H.
C.
Walls of Westinghouse to our Mr. J.
G. Feinstein, dated March 24, 1987 (Identifier AEP-87-205).
This letter makes a correction to Table 2 of the Westinghouse evaluation submitted in AEP:NRC:1024.
Specifically, core fluid volumes for the reference plant and D.
C.
Cook Unit 2 are modified.
The revised values of core fluid volume have been cited in Table II to this attachment, which is presented below.
This letter transmits an analogous evaluation for D.C.
Cook Unit 1 as Exhibit A.
We have transmitted the Unit 1 evaluation in draft form because it is currently undergoing corporate review.
This Unit 1 evaluation will be the subject of a future letter which we anticipate will be transmitted to you in the near future.
4
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Attachment 2 to AEP:NRC:0916Z Page 2
Rather than perform D.
C.
Cook plant-specific analyses to justify operation with the cross-tie valves closed, Westinghouse took the approach of using evaluations based on NOTRUMP analysis results for a reference plant similar in design to the D.
C.
Cook Units.
Additional information related to the decision to use the reference plant approach is provided in Exhibit B to this attachment.
Although the Westinghouse NOTRUMP analyses are not D.
C.
Cook Plant
- specific, use of Westinghouse's NOTRUMP model for the reference plant should be sufficient for examining the impact of different fuel designs on small break LOCA results.
Westinghouse furnished reference plant analyses for both their 17 x 17 standard fuel design and their 15 x 15 "OFA" fuel design.
Table 1 of this Attachment shows that the same component configurations were used in both analyses.
Table 2 presents the plant conditions for both these analyses.
For comparison, Table 2 also give some of those conditions that would represent a core of ANFC 17 x 17 fuel if it were to appear in the reference plant.
For the case where one charging pump injects to four loops (with spilling to containment in one of those loops) and where one high head safety injection pump injects to two loops (with spilling in one of those loops),
the 15 x 15 OFA results in a peak clad temperature of 1427 F.
For the same case for 17 x 17 standard fuel*, the peak clad temperature is estimated at o **
1482 F
Exhibit A of this attachment is the draft description of the Unit 1
- No direct analysis wasperformed for the above cited flow injection configuration for 17 x 17 standard fuel, therefore this estimated result is based on extrapolation from analyses with other flow injections.
- The evaluations considered reference plants of identical power ratings.
(Footnote Continued)
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Attachment 2 to AEP:NRC:0916Z Page 3
15 x 15 OFA work.
The Unit 1 result quoted above appears in Table 4 of Exhibit A.
Exhibit B of this attachment is the actual description of the Unit 2 17 x 17 standard work.
The Unit 2 result quoted above appears toward the top of page 10 in the Westinghouse attachment contained in Exhibit B.
It can be seen that for the difference between the Westinghouse 15 x 15 OFA and 17 x 17 standard
- design, the peak clad temperature difference of 55 F is very small compared to the margin to the acceptance criterion of 2200 F.
The difference between the ANFC 17 x 17 fuel design and the Westinghouse 17 x 17 standard fuel design is not greater than the difference between the two Westinghouse designs presented here.
For illustration, certain parameters for the ANFC design are compared with those for the two Westinghouse designs in
'able 2.
Further comparisons among these designs can be seen in Exhibit B of Attachment 1 to this submittal.
Therefore, any peak clad temperature difference between ANFC and Westinghouse fuel should be no more than on the order of the difference shown for the two Westinghouse
- designs, and the peak clad temperature for Exxon fuel will be similarly much less than the acceptance criterion.
This conclusion is concurred with in the Westinghouse work for Unit 2 (Exhibit B) which states on page 2,
"The difference in fuel pellet outer diameter, fuel rod outer diameter, and fuel rod pitch are small between the Exxon (ANFC) 17 x 17 fuel and the Westinghouse 17 x 17 standard (Footnote Continued)
They differed primarily in the core configuration (15 x 15, versus 17 x 17),
which impacts parameters such as linear hear generation rate.
They also differed slightly in assumed peaking factors:
Unit 2 was analyzed at a peaking factor of 2.40 versus 2.32 for Unit 1.
Westinghouse fuel in both Units is currently licensed for core peaking factors of 2.10. If identical values had been assumed for the peaking factor, the difference between the peak clad temperatures would most likely have been even less than the 55 F
cited above.
Attachment 2 to AEP:NRC:0916Z Page 4
fuel.
Consequently, the effects of fuel parameter differences are expected to have only a small effect on the transient response."
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Attachment 2 to AEP:NRC:0916Z Page 5
Attachment 2
Table 1
Com onent Confi uration Reference 412 Plant Reference 412 Plant with 15 x 15 OFA with 17 x 17 STD VESSEL:
Upper Support Plate Barrel-Baffle Configuration Downcomer Shielding Lower Support Plate Fuel Array Upper Head Spray Flow Percent Top Hat
.Downflow Thermal Shield Curved 15 x 15 OFA 0.21%
Top Hat Downflow Thermal Shield Curved 17 x 17 STD 0.21%
Upper Head Temperature hot hot LOOP COMPONENTS:
Pressurizer Steam Generator Pump Type 1800 ft Model 51 93A; 6000 HP 1800 ft Model 51 93A; 6000 HP
Attachment 2 to AEP:NRC:0916Z Page 6
Attachment 2
Table 2
Plant Conditions Reference 412 Plant Reference 412 Plant Reference 412 Plant 1'5 x 15 OFA 17 x 17 STD ANFC 1 x 17 Licensed Power (MWt)
Licensed Peaking Factor 3338 2.32 3338 2.40 3338 Fuel Volumetric Heat Generation for Total Core (To)al kw/Total ft fuel) 9670.04 9624.0 10902.6 Average Linear Heat Generation (kw/ft)
Total Fluid volume in Core (ft )
(excluding that in fuel assembly guide tubes) 7.033 646.6 5.435 642.9 5.435 679.6 Thermal Design Flow 36950.2 (lbs/sec) 36916.7 Vessel Exit Temperature
( F) 0 608.8 608.0 Vessel Inlet Temperature
( F) 0 544.4 542.2
- No analysis performed, therefore no value generated for comparison.
Exhibit A of Attachment 2
to AEP:NRC:916Z Draft Version of Westinghouse D.
C.
Cook Unit 1 Analysis Regarding Operation with Safety Injection Cross-Tie Valves Closed
'4
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'. DRAFT, There are two high head safety injection (HHSI) pumps in the D.C.Cook Unit 1 design.
Each HHSI pump discharge line splits to deliver flow into two of the four cold legs.
A cross-tie connects the two pump discharge lines enabling one pump to deliver flow to all four of the cold legs.
The design basis small break Loss-of-coolant-accident (LOCA) analyses assume that high head safety injection flow delivery is available through all four lines.
American Electric Power Service Corporation has requested Westinghouse to evaluate the D.C.
Cook Unit 1 Emergency Core Cooling System (ECCS) performance following a small break LOCA for a scenario in which the HHSI cross-tie line is closed during normal full power operation.
The evaluation will serve as the basis for a change to the current LOCA design basis in order to allow full power operation of D.C.
Cook Unit 1 with the HHSI system capable of injection to only two reactor. coolant loops'losure of the cross-tie line results in the flow from one HHSI pump being delivered to only two loops.
This results in a reduction in the amount of total safety injection flow delivery to the RCS during a LOCA event when the single failure of an emergency diesel generator to start following the loss of offsite power is considered.
As a result of the diesel failure, one train of safety injection is lost.
The D.C.
Cook Unit 1 licensing basis LOCA analyses consider both large and small break LOCA events.
The large break LOCA result is not highly dependent on HHSI pump flow capability due to the rapid depressurization to the accumulator actuation pressure (600 psia) and the continued rapid depressurization to the Low Head Safety Injection (LHSI) pump actuation pressure (114.7 psia).
Recovery from a large break LOCA event is governed by the availability of LHSI and accumulator delivery during the reflood phase of the transient,.hence a
reduction in the amount of total HHSI flow delivery will not affect the large break LOCA results.
The small break LOCA result is highly dependent upon charging pump and HHSI pump "flow delivery to the RCSi but is not dependent upon LHSI flow delivery.
Small break LOCA's whi~h result in the highest Peak Cladding Temperatures (PCT's) do not
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'RAFT page 2 of 5 experience primary reactor coolant system depressurization to the LHSI delivery pressure, therefore the small break LOCA analYais considers sa fety injection flow from the charging and HHSZ pumps only.
- Hence, a
change to the design basis in which the HHSI cross-tie line is assumed unavailable requires that only the small break LOCA results be considered.
In order to determine the effect of the cross-tie line closure on the plant response to a small break LOCA, Westinghouse peri'ormed an analysis on a reference plant similar in design to D.C.
Cook Unit l.
The reference four loop plant used to determine the safety injection sensitivity is essentially identical to Cook 1 in vessel design and loop components.
Table 1 provides a comparison of the basic vessel and components design of D.C.Cook Unit 1 with the reference plant design with a 15X15 OFA core.
Table 1 shows that the plant designs
- are, virtually identical except for the upper head bypass flow.
The slight difference in the upper head bypass flow at such low flowrates is expected to have an insignificant effect on plant response to a small break LOCA.
Table 2 provides a comparison of some of the important parameters influencing plant response to a small break LOCA for D.C.Cook Unit 1 and the reference plant design (again with a 15X15 OFA core).
The major differences noted between the plants include the licensed core power and the licensed peaking factor.
Both D.C.Cook Unit 1 and the modified reference plant operate with a 15X15 OFA fuel array.
As noted, the licensed core power level of D.
C.
Cook Unit 1 is lower than the power level assumed in the modified reference plant used for this analysis by 2.74.
The total core power level influences the depth and duration of core uncovery.
That is, the higher the power level, the deeper and longer the core will uncover.
The reactor coolant system response to a small break LOCA, as calculated by the NOTRUMP code, demonstrates a rapid depressurization down to the steam generator secondary relief valve pressure.
This condition represents a
quasi-equilibrium pressure at which the primary system tends to stabilize prior to the venting of steam through the broken pump suction leg loop seal.
Following loop seal venting in the broken loop, core boiloff exceeds the safety injection mass flow rate and a core uncovery transient results.
Therefore, depth and duration of core uncovery prior to reaching the accumulator injection setpoint are dependent upo n not only the initial power level, but also the difference between safety injection flow rate and the core boiloff flow rate.
Since this analysis modelled the 15X15 OPA core, cross tie closed SI flow rates and a higher power level than that licensed for D. C.
Cook Unit 1<
regardless of other differences in the initial
I l
a
'A~
Cg
/ 'f 4'A
ORAFT page 3 of 5 operating conditions, a plant specific analysis for the same break size would result in less limiting results than those presented herein.
A small break LOCA analysis was performed for the reference plant applying the limiting four-inch equivalent diameter cold leg break for the D.C. Cook Unit 1 licensing basis WFLASH analysis.
The analysis assumed.a safety injection flow rate representative of the HHSI flow configuration at D.C.
Cook Unit 1 with the cross tie closed.
Combined with Appendix K minimum SI assumptions, this results in one charging pump available to deliver flow through four lines and one HHSI pump available to deliver flow through two of four lines.
The analysis was performed at 1024 of the reference plant licensed core power assuming a fission product decay heat generation rate of 1.2 times the -1971 ANS Decay Heat values.
The analysis also assumed the loss of offsite power and the single failure of a diesel to start.
Zt was conservatively assumed that all safety injection flow delivered to the broken loop spilled out the break.
As a result of the spilling assumption, safety injection flow from one charging pump is delivered to the three intact RCS loops and one HHSZ pump delivers flow to one of the two remaining RCS loop delivery lines.
The analysis was performed to determine the peak clad temperature using the D.C.
Cook Unit 1 SI flow representative of the cross tie closure.
The reference plant analysis was performed using the Westinghouse NRC approved small break LOCA ECCS evaluation model using the NOTRUMP code as described in WCAP-10054-P-A and WCAP-10079-P-A.
NOTRUMP addresses all of the NRC concerns expressed in NUREG-0611 and meets
.the requirement of NUREG 0737 II.K.3.30.
The analysis resulted in a PCT of 1427'F, thereby illustrating that operation in the flow configuration in which one charging pump is available to deliver flow to four RCS loops and one HHSZ pump is available to deliver flow to only two of four loops, does not violate the requirements of 10 CFR 50.46 and Appendix K.
The reference four loop plant FSAR analysis was performed at 1024 of a licensed core power level of 3338 MWt with a core peaking factor of 2.32 and resulted in a PCT of 1427'F for the limiting four inch break.
Table 5 provides the number of safety injection pumps, the number of lines available to deliver flow to the RCS and the spilling
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DRAFT
"(Includes corrections From Exhibit C)
TABLE 2 PLANT CONDITIONS Actual vs. Modelled PARAMETERS COOK 1
REFERENCE 412 w/D.C.
Cook Unit 1 Core LICENSED POWER (MWt) 3250 3338 LICENSED PEAKING FACTOR 2.10 2.32 FUEL VOLUMETRIC HEAT GENERATION FOR TOTAL CORE (TOTAL KW/TOTAL FT FUEL) 9415.10 9670.04 AVERAGE LINEAR HEAT ENERATION (KW/FT) 6.848 7.033 PEAK LINEAR HEAT GENERATION (KW/FT) 14.380 15.473 CORE FLUID VOLUME (FT ):
Rod Channel Fluid Volume Core Thimble Tube Volume Total Core Fluid Volume 646. 6
- 65. 52 712
~ 12 646. 6 65.52 712 '2 THERMAL DESIGN FLOW (LBS/SEC.)
37666.7 36950.2 VESSEL EXIT TEMPERATURE
( F) 599.3 608.8 VESSEL INLET TEMPERATURE oF) 536.3 544.4
TABLE 3 A
S SS TE S
COOK 1 FSAR ANALY REFERENCE 412 v/D.C. Cook Unit 1 Core SES LICENSED CORE
- POWER, 102%
OF 3411 MWt 3338 MWt PEAK LINEAR HEAT GENERATION 15.811 KW/FT 15.473 KW/FT AVG. LINEAR HEAT GENERATION 7.187 KW/FT 7.033 KW/FT TOTAL CORE PEAKING
- FACTOR, Fg FUEL TYPE 2'2 15X15 OFA WESTINGHOUSE 2.32 15X15 OFA WESTINGHOUSE SMALL BREAK MODEL WFLASH NOTRUMP
L
'h l
lg
. DRAFT TABLE 4 SMA B
K S
S RESULTS COOK UNIT 1 FSAR REFERENCE 412 w/D. C.
Cook Unit 1 Core SMALL BREAK MODEL WFLASH NOTRUMP PEAK CLAD TEMPERATURE 1716 F
1427 F
PEAK CLAD TEMPERATURE CATION (FT.)
- 11. 75
- 12. 0 TIME OF PCT (SEC.)
823 947 REACTOR TRIP (SEC.)
17 F 50 1.46 CORE UNCOVERY TIME (SEC) 413 '
646.5 ACCUM. INJECTION (SEC.)
800.0 880 '
CORE RECOVERY TIME (SEC.)
1310 1143
jX I
t,fr'
Table 5
D.C.
COOK UNIT 1 ORIG.
FSAR ANALYSIS 1
HHSZ ZNJ.
TO 4 LINES 1 CHARGING ZNJ.
TO 4 LINES 1 LINE SPILLS 1 LINE SPILLS REFERENCE 412 vith D:C.
Cook Unit 1 Core 1 HHSI INJ.
TO 2 LINES 1 CHARGING INJ.
TO 4 LINES 1 LINE SPILLS 1 LINE SPILLS
0 C
COOK UNIT 1
(AEP)
NQTRUNP SB LOCA 41Z STO 4-INCH COLD LEG BREAK 15X15 OFA FUEL - X-TIE CLOSED
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NOTRUNP SB LOCA 412 STD 4-INCH COLD LEG BREAK 15X15 OFA FUEL X-TIE CLOSED
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NOTRUMP SB LOCA 412 STD 4-INCH COLD LEG BREAK 15X15 OFA FUEL X-TIE CLOSED
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NOTRUMP SB LOCA 41Z STD 4-INCH COLD LEG BREAK 15X15 OFA FUEL X-TIE CLOSED W ~
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t
0 C
COOK UNIT 1
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NOTRUNP SB LOCA 412 STD 4-INCH COLD LEG BREAK 1'5X15 OFA FUEL X-TIE CLOSED ILJ Il H 4
5 IIIC ISCC 0
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IV 1%
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D C
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(AEP )
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C COOK lJtl! T 1
(AEP)
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- 0. C. Cook Unit i SI Flee
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Exhibit B of Attachment 2
to AEP:NRC:0916Z Letter from M. P. Alexich (I&NECo.)
to H. R. Denton (NRC) Regarding Operation of D.
C.
Cook Unit 2 with Safety Injection Cross-Tie Valves Closed (Previously Submitted in Letter AEP:NRC:1024, Dated March 23, 1987)
Attachment 1 to AEP:NRC:1024 Unit 1 and 2
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Attachment 2 to AEP:NRC:1024 Information Notice 87-01
Attachment 3 to AEP:NRC:1024 I.E. Inspection Report 86042 Unit 1 and Unit 2 LERs
4 ~
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'li gt q ~1i '
September 15, 1986 1150 hrs.
1400 hrs.
Licensee Problem Assessment Group (PAG) determined that this Condition Report involved events which are 10 CFR 50.7g and 10 CFR 50.73 reportable.
10 CFR 50.72 Four-Hour Report Kotification was made.
Root Causes The proximate cause of the event was lack of recognition by licensee personnel that the HPSI cross-tie valves are embodied in the meaning of "flow path" for that system.
Several other factors contributed.
First, the Technical Specifications provide no clear meaning of the term "flow path" as it applies to the SI system.
This is also true for other
- systems, but not all of them and in some cases, the meaning is clear.
- Second, an insufficiently challenging Job Order review and approval process occurred in this case.
This was voluntary maintenance, the leak which was repaired was small, and presented no threat to operability of the valve or other nearby components.
Several NRC licensed Senior Reactor Operators (SROs) had opportunities to question or challenge the proposed maintenance activity as inconsistent with plant MOOE, however, none did so.
The licensee lacked experience in setting up a "clearance" to work on the cross-tie valves.
Neither valve has required maintenance for several years, if ever.
Oespite the fact that the activity involved both a novel and a non-mandatory isolation of safety related equipment, the licensee did not appear to perform anything other than a routine review and approval process.
The licensee performs on average, from 4,000 to 5,000 "clearances" (primarily for maintenance) each year.
Many are repetitive; some are unique.
Applicable administrative control procedures currently provide no special conditions or other criteria for handling uncommon clearances.
The Shift Supervisor (licensed SRO) is ultimately responsible for approving all re~oval from and return to service of safety-related (and other) equipment.
Training considerations are addressed in Paragraph. 7, below.
Event Si nificance Overall safety significance is small.
The margin of safety to fuel damage was reduced but not by more than about 25 percent, for a small spectrum of small break Loss of Coolant Accidents (LOCA's), which are the accidents for which SI pumps provide significant contribution.
4h V
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$ t I>"
The licensee and Westinghouse analyzed the subject I pump/2 cold leg configuration for the 11mftfng break size (4-inch) and determined an average 5.84 percent flow reduction, for ECCS flow to the reactor, would occur fn the critical small break LOCA, as the RCS depressurfzed from 1000 psf to 600 psf.
The analysis credited flow from the two operable centr1fugal charging pumps vfa alternate paths.
At 600 psf, addft1onal flow becomes available from the accumulators.
Previous analyses had established a 21 degree Fahrenheit increase fn peak clad temperature for each percent reduction in flow.
- Thus, a peak clad temperature (PCT) increase of about 123 degrees Fahrenheit could have resulted under these circumstances.
The PCT could then be calculated to be 1791 degrees Fahrenheit with this configuration, rather than the 1668 degrees Fahrenheit originally calculated fn the FSAR for breaks 1n this spectrum.
This constitutes about a 23 percent reduction fn the 532 degree margin of safety to ECCS Rulemakfng PCT of 2200 degrees Fahrenheit.
The event involved a violation of a Technical Specification L1mftfng Condition for Operation for an Emergency Core Cooling System wh1ch resulted fn a reduction of the systems design safety marg1n.
The system configuration that resulted fn the violation was reviewed fn advance by several licensed senior operators and persisted through two shift turnovers without disclosure.
The improper cross-tfe configuration was not identified by control room operators because they dfd not understand safety injection design bas1s due to a lack of specific system design basis tra1nfng.
~Re art$ n Reportf ng requf r ements were met.
A four-hour telephone notff1catf on pursuant to 10 CFR 50.72(b)(2)(111) was made on September 15, 1986 about two hours and 10 minutes after the determinat1on that the cfrcumstances were reportable.
A written Licensee Event Report (LER 316/86026, Rev. 0) pursuant to 10 CFR 50.73(a)(2)(v) was submitted on October 10, 1986 w1thfn the requ1red 30 days.
~Trefntn A review of the training program and dfscuss1ons with shift personnel was performed focusing on the tra1nfng and qualff1cat1ons of personnel involved fn this event.
Licensed personnel were knowledgeable concerning SI system design and concerning administrative control processes for removal and return of safety equipment for maintenance.
The effectiveness of the tra1n1ng process was defic1ent; however, in that factual knowledge about SI system
- des1gn, administrative processes, and the Technical Specff1cat1ons, were not 1ntegrated effect1vely.
The design bases were not adequately understood.
8.
Conclusion' As shown above, the'icensee operated Unit 2 in HODE 1 for about 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br /> with the safety injection system in a condition not considered in the design bases and contrary to Technical Specification requirements.
The primary cause was failure to recognize the significance of the SI cross-tie valves.
This was contributed to by a lack of specificity in Technical Specifications, a lack of challenge in the review and approval process of non-routine clearances and ineffectively integrated training information.
9.
Exit Interview The inspectors met with licensee representatives
{denoted in Paragraph 1) on November 7, 1986 to discussed the scope and findings of the inspection report.
This exit was performed in conjunction with the exit for Inspection Report No. 315/86035; 316/86035.
The inspector asked those in attendance whether they considered any of the items discussed to contain information exempt from disclosure.
No items were identified.
I
Attachment 6 to AEP:NRC:1024 Westinghouse Small-Break LOCA Evaluation for Unit 2
4 t
q'f
ce W
ss e
os e'ACKGRO There are two high head safety injection (HHSI) pumps in the D.C.Cook Unit 2 design.
Each HHSI pump discharge line splits to deliver flow into two of the four cold legs.
A cross-tie connects the two pump discharge lines enabling one pump to deliver flow to all four of the cold legs.
The design basis small break Loss-of-coolant-accident (LOCA) analyses assume that high head safety injection flow delivery is available through all four lines.
American Electric Power Service Corporation has requested Westinghouse to evaluate the D.C.
Cook Unit 2 Emergency Core Cooling System (ECCS).
performance following a small break LOCA for a scenario in which the HHSI cross-tie.line is closed during normal full power operation.
The evaluation will serve as the basis for a change to the current LOCA design basis in order to allow full power operation of D.C.
Cook. Unit 2 with the HHSI system capable of injection to only two reactor coolant loops'losure of the cross-tie line results in the flow from one HHSI pump being delivered to only two loops.
This results in a reduction in the amount of total safety injection flow delivery to the RCS during a LOCA event when the single failure of an emergency diesel generator to start following the loss of offsite power is considered.
As a result of the diesel failure, one train of safety injection is lost.
The D.C. Cook Unit 2 licensing basis LOCA analyses consider both large and small break LOCA events.
The large break LOCA result is not highly dependent on HHSI pump flow capability due to the rapid depressurization to the accumulator actuation pressure (600 psia) and the continued rapid depressurization to the Low Head Safety Injection (LHSI) pump actuation pressure (114.7 psia).
Recovery from a large break LOCA event is governed by the availability of LHSI and accumulator delivery during the reflood phase of the transient, henc~
a reduction in the amount of total HHSI flow delivery will not affect the
P I
Page 2 of 10 large break LOCA results.
The small break LOCA result is highly dependent upon charging pump and HHSI pump flow delivery to the
- RCS, but is not dependent upon LHSI flow delivery.
Small break LOCA's which result in the highest Peak Cladding Temperatures (PCT's) do not experience primary reactor coolant system depressurization to the LHSI delivery pressure, therefore the small break LOCA analysis considers safety injection flow from the charging and HHSI pumps only.
- Hence, a
II change to the design basis in which the HHSI cross-tie line is assumed unavailable requires that only the small break LOCA results be considered.
In order to determine the effect of the cross-tie line closure on the plant.response to a small break LOCA, Westinghouse performed an analysis on a reference plant similar in design to D.C. Cook Unit 2.
. The reference four loop plant used to determine the safety injection sensitivity is essentially identical to Cook 2 in vessel design and loop components.
Table 1 provides a comparison of the basic vessel and components design of D.C.Cook Unit 2 with the reference plant design.
Table 1 shows that the plant designs are virtually identical except for the upper head bypass flow.
The slight difference in the upper head N
bypass flow at such low flowrates is expected to have an insignificant effect on plant response to a small break LOCA.
Table 2 provides a
comparison of some of the important parameters influencing plant response to a small break LOCA for D.C.Cook Unit 2 and the reference plant design.
The major differences noted between the plants include the licensed core power, the licensed peaking factor, and the fuel types'oth D.C.Cook Unit 2 and the reference plant operate with a 17X17 fuel array.
The current cycle of operation at D.C.Cook Unit 2 contains an EXXON fuel core with the exception of one Westinghouse fuel assembly, while the reference plant has a complete 17X17 Westinghouse standard fuel core.
The difference in fuel pellet outer diameter, fuel rod outer diameter, and fuel rod pitch are small between the EXXON 17X17
"'fi k
J(
Page 3 of 10 fuel and the Westinghouse 17X17 standard fuel.
Consequently, the effects of fuel parameter differences are expected to have only a small effect on the transient response.
" The total core power level influences the depth and duration of core uncovery.
The decay heat generation rate, which is a function of the total core power level, determines the rate of core boiloff steam production in conjunction with the safety injection flov rate following a small break LOCA.
The reactor coolant system thermal-hydraulic response to a small break LOCA, as calculated by the NOTRUMP code, is then dependent upon the ratio of the core power to safety injection flow rate.
Higher safety injection flov delivery rates provide more core subcooling to absorb decay heat as veil as providing more mass addition to keep the core covered.
Consequently, the effect of the indifference in the total core power level between the plant designs is diminished due to the influence of the safety injection flov.
The core power to safety injection flow ratio has been determined at a representative pressure of 1114.7, psia which approximately corresponds to the quasi-equilibrium pressure at vhich the. primary system tends to stabilize prior to the venting of the pump suction leg loop seal.:
Following loop seal venting, core boiloff exceeds the safety injection mass flow rate and a core uncovery transient results.
The ratio of core power to the safety injection mass flow rate for the reference plant is 81.814 (MW/ibm/sec).
For D.C.
Cook Unit 2 in the current design basis configuration, one HHSI pump and one charging pump inject through four lines.
The core power to safety injection mass flow rate is 65.345 (MW/ibm/sec).
If a small break LOCA FSAR analysis were performed for D.C.
Cook Unit 2 the results would be less limiting than the result of the FSAR NOTRUMP analysis performed for the reference plant.
Hence the reference four loop plant used in the NOTRUMP analyses is expected to provide a representative thermal hydraulic response to a small break L0CA for variations in the safety injection flow for D.C.
Cook Unit 2.
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Page 4 of 10 MALL A
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A small break LOCA analysis was performed for the reference plant applying the limiting four-inch equivalent diameter cold leg break for the D.C. Cook Unit 2 licensing basis WFLASH analysis.
The analysis assumed a safety injection flow rate representative of the HHSI flow configuration at D.C.
Cook Unit 2 during a recent License Event Report (LER) in which two charging pumps were available to'eliver flow through four lines and one HHSI pump was available to deliver flow
.through two of four lines.
D.C.
Cook Unit 2 issued the LER for operation in Mode 1 (at 804 full power) with one HHSI pump removed from service and the cross tie line closed.
Closure of the cross-tie line did not meet the design basis small break LOCA analysis assumption of HHSI flow delivery to all four loops.
The analysis was performed at 102% of the reference plant licensed core power assuming a fission product decay heat generation rate of 1.2 times the 1971 ANS Decay Heat values.
In order to support the operation of D.C.
Cook Unit 2 with two charging pumps and one HHSI, the loss of offsite power without the single failure of a diesel to start was assumed.
It was conservatively assumed that all safety injection flow delivered to the broken loop spilled out the break.
The analysis was performed to determine the effect on plant response for a change in the SI flow assumption, thereby establishing a safety injection sensitivity for a plant similar in design to D.C. Cook 2.
The reference plant analysis was performed using the Westinghouse NRC approved small break LOCA ECCS evaluation model using the NOTRUMP code as described in WCAP-10054-P-A and WCAP-10079-P-A.
NOTRUMP addresses all of the NRC concerns expressed in NUREG-0611 and meets the requirement of NUREG 0737 II.K.3.30.
The analysis resulted in a PCT of 1132 F, thereby illustrating that operation in the flow configuration in which one charging pump is available to deliver flow to four RCS loops and one HHSI pump is available to deliver flow to only two of four loops, does not violate the requirements of 10 CFR 50.46.
1
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Page 5 of 10 The reference four loop plant FSAR analysis was performed at.1024 of a licensed core power level of 3338 MWt with a core peaking factor of 2.4
.and resulted in a PCT of 1244 F for the limiting four inch break.
The analysis performed to determine the SI sensitivity employed the same assumptions as the reference plant FSAR analysis with the exception of the safety injection flow delivery flowrates.
Table 5
provides the number of safety injection pumps, the number of lines available to deliver flow to the RCS and the spilling assumptions utilized in the analyses and the evaluation.
The safety injection flow vs. pressure curve assumed in the reference four loop plant FSAR analysis is shown in figure 11.
This may be compared to the safety injection flow assumed for the sensitivity analysis.
The safety injection flows for the proposed HHSI ECCS design basis operation with closure of the cross tie connection is also shown in figure 11...
Table 3 provides a comparison of the reference plant analyses assumptions to the D.C.
Cook Unit 2 FSAR analysis assumptions.
The four inch break for the NOTRUMP analysis employing SI flows used in the sensitivity analysis is characterized by a rapid primary side depressurization to a pressure slightly above the steam generator secondary safety valve setpoint.
Steam generator secondary side pressurization to the, safety valve setpoint results from the loss of off-site power assumption.
RCS inventory depletion results in steam venting through the pump suction leg loop seal in the broken loop at approximately 356 seconds.
This permits steam generated in the core to exit through the break, resulting in continued RCS depressurization.
Since the core boiloff rate exceeds the safety injection flow rate, core uncovery results.
Accumulator injection, which reverses the net mass inventory depletion from the RCS, occurs when the RCS pressure decreases to approximately 600 psig.
The safety injection pump performance interval of interest therefore extends from approximately 1000 psig to the accumulator actuation pressure.
Table 4 provides a comparison of the results for the D.C. Cook Unit 2 original FSAR analysis to the reference plant analyses.
The results reported for Cook Unit 2 reflect the original WFLASH analysis specific
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Page 6 of 10 to D.C. Cook Unit 2.
A footnote has been provided which reports the results of the small break analysis performed to determine the effects of a reduction in Safety Injection Flow if the assumed HHSI miniflow is increased from 30 gpm to 60 gpm.
The miniflow analysis was specific to D. C. Cook Unit 1 and was determined to bound operation of D.C.
Cook Unit 2 with the increase in HHSI miniflow. This analysis was used to support the D..C.
Cook Unit 2 license amendment for an increase in HHSI I
miniflow from 30 gpm to 60 gpm.
All references to the D.C.
Cook Unit 2 current design basis SI flows assume the SI available with a HHSI miniflow of 60 gpm.
Figures for the reference plant analyses have been provided to illustrate the influence of the change in safety injection flow rate 'on the pertinent parameters identifying the plant response to the small
.break LOCA.
Figure 1 shows the depressurization transient for the two reference plant analyses for the pressurizer pressure.
Comparison of the two depressurization transients indicates that the change in the safety injection flow rates has very little influence on the system depressurization between the two analyses.
Slight differences are apparent after the time of the pump suction loop seal clearing and are primarily due to the differences in the rate of core boiloff to safety injection flow delivery between the two cases.
Figures 2 and 3
represent the broken loop and intact loop secondary side steam pressure response during the transient respectively.
Comparison of the response for the two analyses shows that they are essentially identical prior to loop seal venting indicating that the amount of primary decay heat removal via the steam generator secondaries is approximately the same in the two cases.
Figure 4 shows the core mixture level which illustrates the major effect of the change in the amount of total safety injection flow available in the two transients.
A comparison of the uncovery transient of the two analyses shows that the reference FSAR case, which has less safety injection flow available than the SI sensitivity case prior to accumulator injection, experiences a deeper core uncovery of slightly longer duration.
Figure 5 represents the pumped safety injection flow for the two cases.
Figure 6 represents
"fi p5
Page 7 of 10 the accumulator flow for the two cases.
The time of initial accumulator injection is slightly delayed in the SZ sensitivity case due to the delay in loop seal steam venting resulting from the increased safety injection mass flow rate.
However, the initial delivery of accumulator water in conjunction with the higher safety injection flows is sufficient to initiate a rapid core recovery and a
higher degree of subcooling which results in additional accumulator delivery'earlier in the SI sensitivity case.
Figure 7 provides a
comparison of the total break flow for the two cases.
In the SI sensitivity case a greater amount of safety injection flow to the intact loop will travel around the downcomer and flow into the broken loop cold leg causing a higher degree of subcooling.
The change in enthalpy resulting from the subcooled.-fluid injection to the broken loop cold leg results in a slightly higher liquid break flow rate from 250 to.375 seconds.
After loop seal clearing, the two cases show essentially identical break flow behavior.
These figures show that the change in the available safety injection water has a negligible impact on the depressurization transient,, the secondary side pressure response and the break flow rate.
The net effect of safety injection flow differences on small break LOCA response is in the depth and duration of the core uncovery transient.
The effect of the reduction in SZ flow on the plant response to a small break LOCA resulting from the proposed design basis change would therefore be to slightly increase the depth and duration of core uncovery.
The increase in PCT resulting from the additional core uncovery can be estimated from the sensitivity obtained from the reference plant analyses.
The peak clad temperature and the peak fuel (pellet) temperature results for the two reference plant analyses are presented in figure 8
and figure 9 respectively.
Zn figure 8 the clad average temperature for the hot rod of the reference plant FSAR analysis is compared to the hot rod clad average temperature of the SZ sensitivity analysis.
The
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Page 8 of 10 results confirm that the case with the higher power to SI flow ratio will result in a more limiting PCT.
The clad average temperature in I
the hottest rod is determined assuming the power in the hottest rod to be equal to the Peak Linear Power (Peaking Factor
Clad heat up calculations specific to D.C.
Cook 2 would result in slightly lower PCT's since the D.C.
Cook 2 Peak Linear Power is limited by the large break LOCA F~.
Figure 9 shows a comparison of the fuel pellet average temperature in the hottest fuel rod for the two cases.
The pellet average temperature for both cases displays the same behavior throughout the transient as the corresponding clad average temperature with slightly higher temperature peaks than the clad average temperature.
Figure 10 provides a comparison of the D.C. Cook Unit 2 current design basis safety injection flow rates to the safety injection flow rates which would be assumed available in the small break LOCA analysis under the proposed design basis change.
The effect of these safety injection flow changes on the small break LOCA peak cladding temperature will be determined from the safety injection flow sensitivity obtained from the reference plant analyses.
Zn Figure 11 the reference plant analyses safety injection delivery rates are compared to the design basis-change safety injection flow rates which consider the flow reduction from the cross tie closure.
An evaluation of the effect of the proposed design basis change safety injection flows on a D.C.
Cook Unit 2 small break LOCA analysis will be performed by establishing the sensitivity of peak clad temperature to variations in safety injection flow from the reference plant analyses.
The effect of the difference in licensed core power between the reference plant and D.C.
Cook Unit 2 has been examined and has been shown to have only a small effect on the small break LOCA PCT.
The total core power level in conjunction with the safety injection flow rate determines the rate of core boiloff steam production following a
~A 0
Page 9 of 10 small break LOCA.
The reactor coolant -system thermal-hydraulic response to a small break LOCA, as calculated by the NOTRUMP code, is then dependent upon the ratio of the core power to safety injection flow rate.
This ratio for the reference plant at the representative pressure of 1114.7 psia for the design basis change SI flows has been determined to be 98.118 (MW/ibm/sec).
At the D.C.
Cook Unit 2 licensed core power level of 3411 MWt for the same safety injection flow, the ratio of power to SZ flow is determined to be 100.26 (MW/ibm/sec).
This indicates that any change in the core uncovery transient between the two cases would be negligible.
Zn addition, the reference case analyses used to determine the sensitivity of PCT to SI flow changes were performed for a core peaking factor (F~) of 2.4.
- Hence, the hot rod clad heat up calculations were performed assuming a peak linear power higher than D.C.
Cook Unit 2.
The effect of the slightly higher
'power to SI flow'ratio for D.C.
Cook Unit 2 is compensated for in the hot rod clad heat up calculations.
The direct sensitivity of PCT to SZ flow changes obtained from the reference plant analyses is. therefore applicable to D.C.
Cook Unit 2 and the evaluated PCT will be representative of a D.C.
Cook Unit 2 small break LOCA FSAR analysis.
A relationship can be established between the change in Peak Clad Temperature and the, change in the total (integrated) safety injection flow delivered from break initiation until accumulator actuation for the two NOTRUMP analyses.
This relationship may then be used to determine the Peak Clad Temperature which would result from a small break LOCA analysis with the safety injection delivery rates available with the HHSI cross tie line closed.
The sensitivity established from the two analyses was determined to be (Delta PCT)/(Delta Total SZ Flow Delivered)
= -0.0162 F/ibm Using this relationship and applying a conservative calculation of the integrated amount of safety injection flow which would be delivered in
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Page 10 of 10 a small break LOCA analysis when the HHSI cross tie is assumed
- closed, an FSAR grade NOTRUMP calculation for D.C.Cook Unit 2 is estimated to result in a PCT of approximately 1482 F.
This estimate is based on the total amount of safety in)ection which would be available during the small break LOCA transient assuming that charging flow delivery from one pump is available to four lines with one line spilling to RCS
- pressure, and that HHSI flow delivery from one pump is available to only two of four lines with one line spilling to RCS pressure.
The integrated amount of safety injection flow for the design basis change evaluation was calculated assuming the depressurization transient of the reference plant analysis.
As shown in figure 1, a change in safety injection flow delivery (in the ranges being considered) has an insignificant effect on the depressurization of the RCS, hence it is.
appropriate to assume the depressurization of the reference plant analysis for the determination of the total safety injection delivered in a small break LOCA analysis applying the proposed design basis change SI flows.
In this estimate the HHSI system cross-tie is assumed to be closed, but both HHSI pumps are assumed to be available to deliver flow to two cold legs.
- However, one train of safety injection flow is assumed to be lost due to the loss of off-site power and the failure of a diesel to start.
h CONCLUS ON Applying the sensitivity to safety injection flow changes developed for a plant similar in design to D.C.
Cook Unit 2, an estimate of the Peak Cladding Temperature which would be obtained from a small break LOCA analysis assuming a reduction in safety injection flow as a result of the HHSI system cross tie valve closure can be obtained..
The estimated PCT assuming HHSI cross tie closure and the single failure of a,.diesel generator to start when all ECCS pumps are available was estimated to be approximately 1482 F.
This indicates there is significant margi.n to the 2200 F limit of 10 CFR 50.46 for the operation of D.C.
Cook Unit 2 at a licensed core power of 3411 MWt with the High Head Safety Injection System cross tie connection closed.
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Com o e ts i
a 'o fe ence 12 VESSEL Upper Support Plate Top Hat Top Hat Barrel Baffle Conf.
Downflow Downflow Downcomer Shielding Thermal Shield Thermal Shield Lower Support Plate Curved Curved Fuel Array 17X17 17X17 Upper Head Spray Flow Percent 0.154 0.21%
Upper Head Temp.
THOT THOT LOOP COMPONENTS:
Pressurizer
~
1800 1800 Steam Generator Model 51 Model 51 Pump Type 93A 6000 Hp 93A 6000 Hp
gl
TABLE 2 ON NS TERS FERENCE 4 12 LICENSED POWER (MWt) 3411 3338 LICENSED PEAKING FACTOR 2'0 2.40 FUEL VOLUMETRIC HEAT GENERATION FOR TOTAL CORE (TOTAL KW/TOTAL FT FUEL) 9834.46 9624.0 AVERAGE LINEAR HEAT ENERATION (KW/FT)
PEAK LINEAR HEAT GENERATION (KW/FT) 5.554
- 11. 663
- 5. 435 13.043 TOTAL FLUID VOLUME IN CORE (FT
)
613. 0 612
~ 9 THERMAL DESIGN FLOW (LBS/SEC.)
37388.9 36916
~ 7 VESSEL EXIT TEMPERATURE
( F) 606.4 608.0 VESSEL INLET TEMPERATURE
( F) 541.3 542 '
TABLE 3 LYS S
ZONS ARAME S
COOK 2 A
ZS REFERENCE 412 ALYSES LICENSED CORE
- POWER, 1024 OF 3391 MWt 3338 MWt PEAK LINEAR HEAT GENERATION 12
~ 81 KW/FT 13.043 KW/FT AVG. LINEAR HEAT GENERATION 5.521 KW/FT 5.435 KW/FT OTAL CORE PEAKING
- FACTOR, Fg FUEL TYPE 2.32 17X17 STD.
WESTINGHOUSE 2.40 17X17 STD WESTINGHOUSE SMALL BREAK M DE WFLASH NOTRUMP
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TABLE 4 L
SU RESULTS COOK UNIT 2 S
REFERENCE 412 REFERENCE 412 SMALL BREAK MODEL WFLASH NOTRUMP NOTRUMP PEAK CLAD TEMPERATURE 1668 F*
1244 F
1132oF "PEAK CLAD TEMPERATURE LOCATION (FT.)
- 11. 25 12 ~ 0
- 12. 0 TIME OF PCT (SEC.)
913 991 1007 EACTOR TRIP (SEC.)
17'2 4'8 4.18 CORE UNCOVERY TIME (SEC) 413
~ 0 704. 0 709.5 ACCUM. INJECTION (SEC.)
875. 0 919. 0 932.0 CORE RECOVERY TIME (SEC.)
1650 1168 1168
- The results of Unit 1 used to license for an PCT of 1716 F.
the small break LOCA analysis specific to D.C.
Cook support the amendment to the D.C. Cook Unit 2 operating increase in HHSI miniflov from 30 gpm to 60 gpm report a
lf sO 1
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Table 5
J ON G
ION COMPARISON m s Avai ab e
es Ava lable L'nes S illin REFERENCE PLANT FSAR ANALYSIS 1 CHARGING 1 HHSI INJ.
TO 4 LINES 1 LINE SPILLS INJ.
TO 4 LINES 1 LINE SPILLS D.C.
COOK UNIT 2 ORIG.
FSAR ANALYSIS 1
CHARGING 1 HHSI ZNJ.
TO 4 LINES INJ.
TO 4 LINES 1 LINE SPILLS 1 LINE SPILLS FERENCE PLANT SI ENSITIVZTY ANALYSIS 2
CHARGING 1 HHSI INJ.
TO 4 LINES INJ.
TO 2 LINES 1 LINE SPILLS 1 LINE SPILLS D.C.
COOK UNIT 2 PROPOSED DESIGN BASIS MODIFICATION 1
CHARGING 1 HHSZ ZNJ.
TO 4 LINES INJ.
TO 2 LINES 1 LINE SPILLS 1 LINE SPILLS
- - - >> - Reference Four-Loop NOTRUMP SI Sensitivity Reference Pour-Loop FSAR NOTRUMP
>> ~IKW
~lK~
>> ~I%M
~ICM
~Mg No IIIC ISKCI
~IC'H el REFERENCE i-LOOP NOTRUMP FSAR SI FLOMS vs.
SI SENSITIVITY FLOWS Figure 1
- - - - - Reference Pour-Loop NOTRUMP SI Sensitivity Reference Pour-Loop FS'AR NOTRUMP
~ IICC H
~IIC'H c
~IÃC~H 4
4No IN+
IIIC ISCC I
~IC~H
~IÃ'M REFERENCE g Lppp NpTRUMP FS~ SI PZg)gS vs.
SI SENSITIVITY F Figure 2
1
- - - - - Reference Pour-Loop NOTRUHP SI Sensitivity Reference Four-Loop FSAR NOTRUMP
~IIL+H
~l<<C'H
~IC+H e.
<<e.
OO JNo
~
&e 404
~
~
lllC ISCCI
~IX~H REFERENCE i-IDOP NOTRUMP PSAR SI FZDWS vs+
SI SENSIT?VITY FLOWS Figure 3
- - - - - Reference Four-Loop NOTRUMP SI Sensitivity Reference Four-Loop FSAR NOTRUMP 4o TOP OF HOT LEG I
BOTTOM OF HOT LEG TOP OF CORE ONE TllC ISCCI
~IC' l
REFERENCE 4-LOOP NOTRUMP FSAR SI FQ)WS vs.
SI SENSITIVITY FIDWS Figure i
- - - - - Reference Four-Loop NOTRUMP SI Sensitivtiy.
Reference Four-Loop FSAR NOTRUMP' s.
~a
%st B
Roe
$ s.
CMo 1tlC IKCI
~I
~N
~
~H REFERENCE i-LOOP NOTRUMP FSAR SI FLOWS vs.
SI SENSITIVITY FLOWS I
Figure 5
iVy AJ
- - - - - Reference Four-Loop NOTRUMp SI Sensitivity Reference Four-Loop FSAR NOTRUMP
)I I
~.
4&+
TllC ISCC I
~I ~N ol 'N REFERENCE 4-LOOp NOTRUNp FSAR SZ pLOWS vs, SI SENSITIVITY FLOWS Figure 6
AE'I
~
- - - - - Reference Four-Loop NOTRUMP SI fiENSITIVITY FLOWS Reference Pour-Loop FSAR NOTRUMF
~%44
~'N
~ ~I%M
~fC~
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~o
~IC~H ol
~N REFZREgt:E g-ZDOP NPTRUNP FBAR fiI PAWS VSo SI SENSITIVITY FLOWS Figure 7
4%e
>> - - - - Reference Pour-Loop NOTRUMP SI Sensitivity Reference Pour-Loop FSAR NOTRUMP
~ 0COOo LI oo5
~ 1$00o
~J~ INo
~.
CSOo 700o 7Qo COOo ISOo COo
%0o IOCOo IOQo IIOOo IIQo IR4o II% IMCI REFERENCE 4 LOOP NOTRUMP PSAR SI FLOWS vs ~
SI SENSITIVITY PLOWS Figure 8
- - - - - Reference Four-Loop NOTRUMP SI Sensitivity Reference Four-Loop FSAR NOTRUMP ISOOo Lo g I%So C
Mo IIN+ I ISIS I
~
Ms ISO'So
%le IMo l 88o TlK ISCCI REFERENCE 4-LOOP NOTRUMP FSAR SI FQ)WS vs, SI SENSITIVITY-FLOWS Figure 9
110.0 Figure io D. C.
COOK UNIT 2 SI FLOMS OESI6N BASIS SI FLOM VS X-TIE CLOSED SI
NIS FSlR SI XTIE CLSD SI 100.0 90.0 80.0 70.0 60.0 50.0 40.0
(
1 30.0 20.00,0 200.0 400.0 600.0 800.0 1000.0 1200.0 1400.0 PRESSURE tPSIA)
i'.0 Figure ii SAFETY INJECTION FLOM CO%'ARISON REF.
ANALYSIS SI VS. X-TIE CLOSE SI
CLOSED X-TEE.REF.FSAS S1
REF.SI SENS ii0.0 i40.0 90.0
~ ~ ~ ~
80.0 70.0 60.0 50.0 40.0
- I i
I X
~
30.0 RO.O0.0 200.0 400.0 600.0 800.0 i000.0 iROO.O i400.0 PRESSURE lml)
Exhibit C of Attachment 2
to AEP:NRC:0916Z Westinghouse Electric Corp.
Letter AEP-87-200 Related to Our Submittal AEP:NRC:1024
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Westinghouse Electric Corporation Power Systems Nuclear technology Systems Oivision Box 355 Prrrsburgh Pennsyivanra 15230 0355 Mr. J.
G. Feinstein, Manager Nuclear Safety
& Licensing Section American Electric Power Service Corporation One Riverside Plaza
- Columbus, Ohio 43216 AEP-87-205 March 24, 1987 NS-OPLS-OPL-II-87-062 REF:
AEP-87-158 dated March 5, 1987 Attention:
W. G. Harvey AMERICAN ELECTRIC POWER SERVICE CORPORATION D. C.
COOK UNIT ECCS PERFORMANCE/CROSS-TIE CLOSURE
Dear Mr. Feinstein:
The above reference transmitted the final reports which examined both the effect that closure of the High Head Safety Injection (HHSI) cross-tie valve would have on the Emergency Core Cooling System (ECCS) performance in the event of a small break LOCA (Unit 2) as well as the effect on the ECCS following a Large Break LOCA in which the HHSI or Residual Heat Removal cross-tie was closed during normal full power operation (Unit l).
As a result of telecons between Westinghouse and AEP (Ted Zimmerman) on March 23rd and March 24th it was identified that there was an error in the values reported for the core fluid volume for the reference plant and Cook Unit 2.
The core fluid volume values have been corrected in the attached tables and should be incorporated into the final cross-tie closure reports prior to transmittal to the NRC.
Please find attached a revised Table 2 for both the D. C.
Cook Unit 1 and Unit 2 cross-tie closure reports.
If you have any questions, please do not hesitate to contact us.
D. L. Cecchett/dmr Attachment Very truly yours,
~ c-H.
C. Walls, Project Manager Mid-America Area U. S. Nuclear Projects
ATTACHMENT T0 AEP-87-205 TABLE 2 LANT CONDITIONS Actual vs. Modelled PARAMETERS
~COOK REFERENCE 412 v/D.C. Cook Unit 1 Core LICENSED POWER (MWt) 3250 3338 LICENSED PEAKING FACTOR 2.10 2'2 FUEL VOLUMETRIC HEAT GENERATION FOR TOTAL CORE (TOTAL KW/TOTAL FT FUEL) 9415
~ 10 9670.04 AVERAGE LINEAR HEAT NERATION (KW/FT) 6.848 7.033 PEAK LINEAR HEAT GENERATION (KW/FT) 14.380 15.473 CORE FLUID VOLUME (FT ):
Rod Channel Fluid Volume Core Thimble Tube Volume Total Core Fluid Volume 646.
6'5'2 712 '2 646. 6 65'2 712.12 THERMAL DESIGN FLOW (LBS/SEC.)
37666.7 36950.2 VESSEL EXIT TEMPERATURE
( F) 599.3 608.8 SSEL INLET TEMPERATURE F) 536 3
544.4
ATTACHMENT TO AEP-87-205 TABLE 2 PLANT CONDITIONS PARAMETERS COOK 2 REFERENCE 412 LICENSED POWER (MWt) 3411 3338 LICENSED PEAKING FACTOR 2.10 2.40 FUEL VOLUMETRIC HEAT GENERATION FOR TOTAL CORE (TOTAL KW/TOTAL FT FUEL) 9834 '6 9624.0 AVERAGE LINEAR HEAT GENERATION (KW/FT)
PEAK LINEAR HEAT GENERATION (KW/FT) 5.554 11.663 5.435 13.043 TOTAL FLUID VOLUME IN CORE*
(FT
)
643.0 642.9 THERMAL DESIGN FLOW (LBS/SEC.)
37388.9 36916.7 VESSEL EXIT TEMPERATURE
( F) 606 4
608.0 VESSEL INLET TEMPERATURE
( F) 541.3 542.2
- Core Fluid Volume Does Not Include Core Thimble Tube Volume
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