ML20040A600
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8[,t g NUCLEAR REGULATORY COMMisslON WASHlf4GTON, D. C. 20555
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., oj EIE 4 I970 Docket No. 50-561 MEMORANDUM FOR:
D. B. Vassallo, Assistant Director for LWR's, DPM FROM:
D. F. Ross, Jr., Assistant Director for Reactor Safety, DSS
SUBJECT:
SER INPUT FOR BSAR-205 t
Plant Name:
BSAR-205 Docket Number:
50-561 Licensing Stage:
PDA Milestone Number:
27-01 Responsible Branch LWR-3 and Project Manager:
T. Cox Requested Completion Date:
February 24, 1978 Review Status:
Complete
References:
1.
Report to the Advisory Comittee on Reactor Safeguards for BSAR-205, dated July 8, 1977.
2.
Memorandum from D. F. Ross, Jr., to D. B. Vassallo dated November 8, 1977.
3.
Memorandum from T. Ippolite to T. Cox of March 16, 1978 Reference 1 described in SSAR-205 open issues in the Reactor Systems Branch area.
Reference 2 transmitted our review and closed out all but one area, Open Issue Number 9, Emergency Core Cooling System.
Our Emergency Core Cooling System evaluation is enclosed as a revision to reference 1 (enclosure 1). We consider this item resolved.
We have comple-ted our review of the Amendment 21 submittal on intersystem leakage.
Our review of this material is attached as a revision to Section 5.2.5 of reference 1.
Additional revisions are enclosed tc address:
1.
Section 6.3.2 - changes the normally aligned power supply for the spare HPI pump for makeup line break consideracions, e
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Section 5.4.3 - adds a requirement ;:ertaining to check valve leak tests g
as required by Section XI of the ASME Code.
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Section 5.4.3 - adds the recently implemented staff position to cooldown
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and depressurize the plant using only safety grade systems assuming g$g loss of offsite power and a single active failure.
We will require a g
committment to this positicn prior to granting a PDA.
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Contact:
ma.r S. Newberry, NRR x27241
APR 4 578 D. B. Vassallo
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These revisions are contained in enclosure 2.
Amendment 22 to BSAR-205 consisted of additional information pertaining to the experimental results obtained at the B&W Vessel Model Flow Test Facility.
Our acceptance of the BSAR-205 Chapter 15 transient and accident analyses is contingent upon the review and acceptance of this material -
by AB. Also, a recent B&W submittal in Amendment 18 to BSAR-205 altered the ESFAS signals to the main steam isolation valves.
Our previous acceptance of the BSAR-205 accident analyses is now contingent l
upon the review and acceptance of this protection feature as discussed in reference 3.
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- 0. F.
ss, Jr., Assistan Director for Reactor Safety Division of Systems Safety
Enclosures:
1.
ECCS Evaluation 2.
All other Ref. 1 Rev.
cc:
S. Hanauer R. Tedesco s
R. Mattson Z. Rosztoczy D. Ross
- 0. Parr T. Cox T. Novak G. Mazetis S. Israel S.Newberryh i
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C=en Easue Number 9 - D:ergency Core Cooling Syatem Revise the laat sentence on page 6-t5, Section 6.3.4 to read,
... additional analyses were requested..."
Remove ihe tuo sentences in the firs: paragraph at the top of page 6-28 and inser: the following evaluation:
Babcock and Wilcox submitted the additional calculations (reference 1) to show that the results reported in BAW-10102 remain valid and conserv-ative.
The additional analysis evaluated three large breaks (Double Ended Pump Discharge, C = 1.0, 0.8. 0.6) and were performed using the 0
c Babcock and Wilcox evaluation model as described in BAW-10104, Revision 3.
References 2, 3, and 4 provide the staff evaluation of The submitted calculations included Wee uhreWewed chihge.groval.
this model and the changes made since the models initial aps is'follows:
(1)
The peak linear pcwer level utilized in the CRAFT 2 code was reduced frem 18 kw/ft to 16.8.kw/ft.
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j (2) A 0.25 psi differential penalty was added to the cold leg pipe to account for the effect of steam-water interaction in the cold leg pipes due to High pressure Injection.
s (3)
A mcmentum pressure drop from the downcomer to the break was in-cluded in the REFLOOD code.
The staff requested additional information to address each of these modifications (reference 5) and Babcock and Wilco. provided the additional information in response to this request in reference 6.
The peak linear power level input change to 16.8 kw/ft in CRAFT 2 was incorporated to remove an unnecessary conservatism in the analysis.
Since the maximum allcwable peak linear heat rate, as calculated in BAW-10102, is 16.8 kw/ft, this change is appropriate and acceptable.
The O.25 psi differential penalty was previously required by the staff (reference 2) and is also acceptable.
The addition of the momentum pressure drop effects in the reflood calculations for flow frem the down-ccmer to the break in the cold leg were included to obtain a more appropriate break flow.
The staff concludes in reference 7 that the inclusion of this change constitutes a model change, is acceptable for licensing calculations, and that previous calculations performed without this modification are conservative and therefore acceptable.
The resultant decrease in the calculated peak cladding temperature by including the pressure drop is approximately 30*F.
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2 This analysis confirms that the worst break remains the DEPO 8.55 ft (C =1.0) break.
The PCT at the core midplane for this break is 2114*F, n
whTch is 12*F less than that previously calculated and reported in BAW-10102.
Babcock and Wilcox also provided justification to show that the sensitivity studies and break spectrum shape in BAW-10102 were not altered by the above three changes.
1 We conclude that the calculation provided by Babcock and Wilcox shows that BAW-10102 is an appropriate and conservative reference for BSAR-205 to show full compliance with 10 CFR 50.46.
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References
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Ltr. from J. Taylor to S. Varga, "Large Break ECCS Evaluation of B&W's 205 FA NSS Using the August 1977 ECCS Evaluation Model (BAW-10104, Rev. 3), dated September 30, 1977.
- 2..Ltr. from A. Schwencer to K. Suhrke dated January 8,1976.
3.
Ltr. from S. Varga to K. Suhrke, "B&W ECCS Evaluation Model" dated February 18, 1977.
4.
Ltr. from S. Varga to J. Taylor, " Error Corrections in the CRAFT 2 code",
dated May 13, 1977.
5.
Ltr. from S. Varga to J. Taylor, " Review of Topical Report BAW-10104, Rev. 3" dated Ncvember 23, 1977.
6.
Ltr. from J. Taylor to S. Varga dated December 7, 1977.
7.
Memorandum for D. B. Vassallo and K. R. Goller.from D. F. Ross, Jr.,
." Babcock and Wilcox Reflood Model Changes", dated February 8,1978.
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Revise Section S.2.5 to read as folicus; Reactor coolant pressure boundary leakage detection methods with the ex-ception of a portion of intersystem leakage detection, have been identified as outside the scope of BSAR-205.
Babcock and Wilcox has described the methods to detect leakage from the reactor coolant system to the Decay Heat Removal, Makeup and Purification, and Core Flood systems.
Leakage to the Decay Haat Removal System will be detected by the lifting of the system safety valves which discharge to the radwaste system.
This system is in the referencing applicant's scope and will be reviewed on a plant-specific basis.
Leakage into the Makeup and Purification System frcm sources other than the normal letdcwn flow path would be detected by an increase in the normally required makeup flow.
BSAR 205 requires that leakage to the ccmponent cooling water system and main steam system be monitored,by systems supplied by the referencing applicant.
We will review these systems on a plant-specific basis.
The leakage detection systems and procedures utilized by the applicant must be capable of maintaining intersystem leakage within the limits assumed in the BSAR-205 accident analysis.
The i
systems provided in the BSAR-205 NSSS and the associated interface require-ments provided for the detection of intersystem leakage meet the requirements l
of Regulatory Guide 1.45 and satisfy the requirements of General Design Criteria 30.
The diverse methods which must be employed for all other l
leakage detection to satisfy these requirements must be provided in the Preliminary Safety Analysis Report of each application which references BSAR-205.
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Chance. Section 5.4.3, Decau 2ect Removal System, as follous:
Delete the last sentence in the ne: to last paragraph and the last paragraph.
Insert the foiicuing paragraph:
The staff requires that Babcock and Wilcox show that the plant can be cooled and depressurized using only safety grade equipment, assuming loss of offsite power and a single active failure to the level required for decay heat removal system actuation. A committment to the follo"ing requirements must be provided on BSAR-205 prior to granting the PDA.
Particular system modifications will be reviewed as a post-PDA item.
1.
Provide safety-grade steam generator dump valves, operators, air and power supplies which meet the single failure criterion."
2.
Provide the capability to cooldown to cold shutdown assuming the most limiting single failure in less than 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> or show that manual actions inside or outside containment or return to hot standby until the manual actions or maintenance can be performed to correct the failure provides an acceptable alternative.
3.
Provide the capability to.depressurize the reactor coolant system with only safety grade systems assuming a single failure, or show that manual actions inside or outside containment or remaining at hot standby until manual actions or repairs are complete provides an acceptable. alternative.
4.
Provide the cacability for bo' ration with only safety grade systems s
assuming a single failure or show that manual actions inside or outside g containment or remaining at hot standby until manual action or repairs 3,
are completed provides an acceptable alternative.
5.
Conduct tests to study'the mixing of the added borated water and the cooldown under natural circulation conditions with and without a single failure of a steam generator atmospheric dump valve.
These tests and analyses will be used to obtain information on cooldown times and the corresponding AFW requirements.
6.
Ccmmit to providing specific procedures for cooling down using natural circulation and submit a summary of these procedures.
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Provide or require a seismic Category I AFW supply for at least 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> at Hot Shutdown plus cooldown to the DHR system cut-in based on the largest time (for only onsite or offsite power and assuming the worst single failure), or show that an adequate alternate seismic Category I source will be available.
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Add as laac par: graph to the acme section 5.4.3 The staff requires that the plant have the capability to individually leak
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test certain check valves 'n the DHR system and ECCS as required by Section XI of the ASME Code.
Applicants referencing BSAR-205 must provide the capability i
to test those check valves designated as Category A in accordance with Section XI. The capability must be provided for all Category A values 1
and must include, but not limited to, the following valves.
Valve Location CFMW 10 A, B LPI & CFT Discharge CFMW 9 A, B CFT Discharge DHMV 1 A, B LPI Discharge MV 30 Auxiliary Pressurizer Spray MV 28 Auxiliary Pressurizer Spray The staff will review the referencing applicants design at the Construction Pennit Stage for the ccmmittment to leak check these valves.
The testing l
frequency and specific procedures will be reviewed during the final design stage.
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s Revise the second sentence, 6th :c m rcch in Section G.3.2 on I
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?:ce 6-!.2 to read:
1 A third high pressure injection pump will be provided as an installed spare with the electric power to be provided by the same power supply as the l
pump being used for normal makeup.
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