ML20235C072

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Forwards Request for Addl Info Re Amends 3,4 & 5 to PSAR, Per 690212-13 Meetings &
ML20235C072
Person / Time
Site: Brunswick, 05000000
Issue date: 03/05/1969
From: Morris P
US ATOMIC ENERGY COMMISSION (AEC)
To: Colby P
CAROLINA POWER & LIGHT CO.
Shared Package
ML20235B311 List: ... further results
References
FOIA-87-111 NUDOCS 8709240340
Download: ML20235C072 (8)
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Oistributica:

JC PDR Docket File (2) r DRL Reading MR 5 1969 RPB-5 Reading P. A. Morris, DRL

,, Assistant Directors, DRL Br. Chiefs, DRL Docket Nos. 50-324' S. M. Kari (2) 50-325 W. Butler Mr. P. S. Colby Senior Vice President Operating & Engineering Group Carolina Power & Light Company 336 Payetteville Street Raleigh, North Carolina 27602 Dear Mr. Colby On February 12-13, 1969, we met with representatives of your firm to discuss your Amendments 3, 4, and 5 to the Preliminary Safety Analysis Report for the Brunswick Steam. Electric Plant, Units 1 and 2. We discussed the material submitted in these three Amendments; and as indicated in the meeting, we concluded that some of your answers to questions were not responsive and that additional information would be required in about fif ty of the areas identified in our letter of Decacher 12, 1968.

We have included in Part A of the enclosure a summary of the information '

needed with respect to the nuclear steam supply system together with a eummary of.our comments given et the meeting as to why we consider the answere unresponsive. We understand, based upon discussions of Febru-  !

ary 12-13,1969, that you will amend your response to the remainics axeas as if.sted in Part B of the enclosure, which have not been restated. Inf ormation on a few additional areas is also requested in i Part C of the enclosure.

Please contact us if you desire additional discussion or clarification of the material requested.

Sincerely, 1

I edstent sicus by Potet A. Morris l l

Peter A. Morris, Director l Division of Reactor Licensing  ;

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Enclosures:

Parts A, B. & C )

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8709240340 870921 PDR FOIA MENZ87-111 PDR '

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  • Request for Additional Information. -

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Carolina Power and Light Company l (Docket Nos. 50-324 & 50-325) i 1

This request for additional information is made up of three partst . Parts A l 3..

and B relate to the questions. transmitted _to the Applicant on December 12, 1968. Part A provides additional clarifying counsents on portions of the j

. original request.for information. Part B is a listing by number of those I original questions for which the applicant has agreed to provide information >

in addition to that contained in Amendment Nos. 3 and 4 Part C. includes

' questions that were not previously asked of this applicant, but for which .

responses are needed to complete'the application. You may wish to incor-porate your response to the questions in Part C by reference. The question f'

numbers relate to the numbering system used in our December 12, 1968, request for information.- .

1 Part A' I

This part of our request for additional information includes a restatement of the original questions transmitted in our letter of December 12, 1968, followed by additional clarifying remarks.

3.1 PLEASE PROVIDE THE FOLLOWING ADDITIONAL INFORMATION WITH RESPECT TO REACTOR DESIGN PARAMETERS:

3.1.3 THE ANTICIPATED MECHANISMS AND CONSEQUENCES OF FUEL FAILURE FROM EXCESSIVE BURNUP IF IT WERE TO OCCUR DURING NORMAL PUIL 3 POW R OPERATION:

Tour response to Question 3.1.3 is considered incomplete. You indicated that fuel failure from' conditions of excessive burnup ". . . would result in some release of radioactive fission products from the rod or ads involved."

We require additional information that would identify the extent to which this type of fuel failure would have to progress before the fuel failure ,

1 detection devices would provide an indication to the power plant operator.

3.1.6 1RE ANALYTICAL MODEL, ANALYSIS METHOD, AND CORRESPONDING UNCERTAINTIES ASSOCIATED WITH FREDICTION OF THE MAXIMUM FAST NEUTRON FLUENCE AT THE REACTOR VESSEL INNER SURFACE; AND 3.1.7 THE MID-FLANE RADIAL DISTRIBUTION OF FAST NEUTRON FLUX FROM THE CORE SURFACE TO THE REACTOR YESSEL WALL AVERACED OVER THE ANTICIPATED REACTOR VESSEL LIFE; 1

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We are concerned that your predicted f ast neutron fluence might be low by a significant amount. Please augment your response by providing the anticipated radial and axial fission density profile across the core, averaged over the power plant's life, and the removal cross sections used for both the steel and the water regions of the model. Please comment on the ". . . uncertainties associated with prediction of the maximum fast neutron fluence at the reactor vessel inner surface" and state how azin:uthal asymmetries are considered in the analyais.

3 3.2 PLEASE PROVIDE A DISCUSSION OF THE FOLLOWING SUBJECTS RELATED TO CORE THERMAL AND HYDRAULIC FEATURES.

3.2.2 THE POTENTIAL CONSEQUENCES OF A SINGLE FUEL PIN OPERATING FOR EXTENDED PERIODS BEYOND THE CRITICAL HEAT FLUX (CHF) DURING NORMAL FULL POWER OPERATION INCLUDING THE EXPECTED EFFECTS IN AND AROUND THOSE FUEL PELLETS LOCATED IN THE VICINITY OF THE CHF SURFACES;. --

Please augment your response to this question by considering, in much' greater detail, the mechanisms of fuel failure if the postulated conditions were imposed at various times during the fuel cycle and whether the failure can propagate to neighboring fuel elements. - -

4.7 PLEASE COMPARE YOUR PROPOSED DESIGN CRITERIA WITH THE CRITEkIA F0it "

EMERGENCY AND FAULT CONDITIONS CIVEN IN THE SUMMER 1968 ADDENDA  :-

(PARAGRAPHS N-417.10 and N-417.11) TO SECTION III 0F THE ASME BOILER L T l AND PRESSURE VESSEL CODE. DISCUSS ANY DEVIATIONS BETWEEN YOUR CRITERIA .'

AND THOSE ADOPTED BY THE ASME AS CITED ABOVE. .-

l We require an assesemant of the degree to which your criteria given on Tables V-4-1 and V-4-2 deviate from those of the above cited Addenda and your .

justification for each deviation. -

5.1.11 PLEASE PROVIDE INFORMATION ON THE DYNAMIC LOADING CRITERIA AND STRESS AND DEFORMATION CHARACTERISTICS EMPLOYED IN THE DESIGN OF CRITICAL CONTROLS AND INSTRUMENTATION FOR EARTHQUAKE CONDITIONS.

You indicated in your response that you wili ". . . confirm that there is no loss of function due to the design accelerations and interaction effects between the equipment anchorage and its foundation."

Please augment this response with a statement on how it will be con-firmed and the criteria that will be used in this confirmation.

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6.2 PROVIDE YOUR EXPERIMENTAL AND/OR THEORETICAL ASSESSMENT OF THE PROBLEM OF HYDROGEN ACCUMULATION FOLLOWING A LOSS-OF-COOLANT ACCIDENT. ALSO INCLUDE.A DESCRIPTION OF THE METHOD PROPOSED TO CONTROL THE POST-ACCIDENT HYDROGEN ACCUMULATION IN THE CONTAINMENT.

During recent safety reviews of nuclear f acilities, we have found that there may be a potential problem associated with the radiclytic decomposition of water following a loss-of-coolant accident. The Advisory l

Committee on Reactor. Safeguards also addressed this problem in its Decem-bar 12, 1968, letter to the Chairman on Oyster Creek Huclear Power Plant Unit No. 1. This letter included the.following paragraph:

" Studies are continuing on the possible effects of radiolysis of water in the unlikely event of r loss-of-coolant accident. These studies should be evaluated by the Regulatory Staff and appropriate measures taken as deemed necessary."

In order that we may evaluate this potential problem for the Brunswick Steam Electric Plant and complete our review of your application for a Con-struction Permit, please respond to tha following questions:

6.2.1 Provide a summary of the results of applicable analytical and experimental work completed to date on radiolytic decomposition of water, and indicats areas which are not yet complete.

l 6.2.2 Diseums the R&D effort contemplated which would provide infor-mation on the areas which are not yet completed. Indicate the time schedule of this work.

6.2.3 Based on presently available or anticipated information, provide an evaluation of the safety significance of radiolysis producte in the Brunswick containment vessel, af ter an accident. Include buildup of radiolysis products as a function of time, the potential 'for and consequences of recombination, and the ef fects of containment inerting.

As a result of this R&D effort you may determine that post-acci' dent radiolysis will require that you consider the installation of equipment to mitigate the presence of radiolysis products. Should this be the case, we would be interested in any preliminary thoughts you may have on the criterion which will be used to determine whether such equipment will be required.

Further, your preliminary thoughts on the design of the equipment which may be required would be of interest.

e i f .

4-a We recognise that some of_the information requested may be available in the.public record in the context of our regulatory review of similar features of other facilities. If.such is the case, you may wish to'incor-porate the information by reference in your application.

6.4 DISCUSS WHETRER THE ECCS DESIGN HAS INDEPENDENT AND REDUNDANT FEAWRES FRG( THE HEAT SOURCE TO THE REAT SINK AND THE ERTENT THAT ANT SINGLE -

FAILURE CAN BE ACCOMMODATED WITHOUT LOSS OF FUNCTION.

Please amplify ~your response to include consideration of a passive ecaponent failure, such as a pipe or valve upstream of the core spray pumps.

Discuss the consequences.in terms of pump NFSH requirements as well as the potential offsite doses if the possive component failure were to occur af ter a design basis' accident.

7.5 DISCUSS THE PLANT'S CAPABILITY FOR DETECTION OF FUEL FAILURES. THIS DISCUSSION SHOULD INCLUDE THE DETECTION TIME AS A FUNCTION OF FUEL FAILURE SEVERITT.

We wish to evaluate. the sensitivity of your system for the detection of failed fuel. Flesse quantify your statement that "A small number of grossly failed feel rods are almost instantaneously detected as are a large number of pin-hole failures in many rods."

7.7 WE UNDERSTAND THAT THE RBM SYSTEM IS BTPASSED DURING THE STARTUF MODE OF OPERATION, IF THIS IS CORhhCT, WHAT PROTECTION IS THERE ACAINST 1 ROD WITHDRAWAL EXCURSIONS ORICIKATING DURING THE STARTUF MODEt BY A )

ROD WITHDRAWAL EXCURSION WE MEAN THE SIMULTANEOUS WITHDRAWAL OF TWO I RODS (AS A RESULT OF A SINGLE FAILURE IN M E CONTROL STSTEM) TERMINATED l' BT ACTION OF THE APRM SYSTEM AND/0R THE R3M SYSTEM, In your response to the above, you stated that the RBM system is not j used during the startup mode. Please clarify the phrase "is not used."

J For exemple, does it imply a bypasat If so, what specific instrument and/or j circuits accomplish the bypaset '

7.10 FACE VIl-7-2, PARAGRAPH 7.1.2.3.1 IMPLIES THAT THE APRM SCRAM CIRCUITS ARE DEFEATED UNDER CERTAIN CIRCUMSTANCES. PLEASE CLARIFY.

We require a verification as to whether the AFRM system is bypassed  ;

during the startup mode.

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I 10.2 DISCUSS 1HE POTENTIAL FOR AND CONSEQUENCES OF INADVERTENTLY DROPPING LOADED FUEL CASK FROM ITS HIGHEST ELEVATION IN THE SECONDARY CONTAINMENT BUILDING ON TO A CONCRETE FAD. ,

I Your response to this question is incomplete. Please consider the off- {

site doses that might result from such an event, and_if necessary, the l character of procedures and equipment needed to mitigate the consequences.  !

I 14.5 AN ASSUMPTION OF 100% CARRYOVER IS STATED TO BE CONSERVATIVE IN THE BLOWDOWN ANALYSIS. WHAT IS THE EFFECT OF CARRYOVER ON PEAK DRTWELL j 4

FRESSURE AS PREDICTED ANALYTICALLY AND HAS TRIS EFFECT BEEN MFASURED

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EXPERIMENTALLY-In order that we may evaluate your analytical method, please amplify your response to the above question by the following:

14.5.1 Provide pressure vs. time data for the suppression charsber and drywell for each of the measured and calculated points that appear on Figure III-3-1 of Amendment 11, Oyster Creek Nuclear Fower Plant, Unit 1, Docket No. 50-219; I

14.5.2 Provide all the particulars on the measured data, e.g., per- l cent carryover, initial conditions (temperature, relative humidity, and pre purge) for the suppression chamber and drywell, the pressure, mass, and average enthsipy of the water in the reactor vessel before and after the blowdown, and the time for blowdown:

  1. 14.5,3 Estimeta, for the calculated points referred to in 14.5.1, above, the factor by which the calculated peak pressure is expected to exceed the setual pa6k pressure for each of the conservative assumptions made in the calculations. These assumptions included no condensation in the drywell,1001 carryover, and Moody's blowdown model; and 14.5.4 As in 14.5.2, provide the experimental particulars for Test

' Numbers B-32. B-40, H-10, H-16,11-35, and H-45.

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Zart B This section of the enclosure identifies the questions originally transmitted to the applicant on December 12, 1968, for which the applicant agreed to pro- )

vide additional information in our discussion of February 12-13, 1969.

1. 1.1 13. 4.4 25. 5.2.15 i
2. 1.2.1 14. 5.1.1 26. 5.4.4 l q
3. 1.3.5 15. 5.1.3 27. 5.4.5
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4. 1.4.3 16. 5.1.6 28. 7.4 '
5. 1.5 17. 5.1.7 29. 8.1 i j
6. 1.7 18. 5.2.1 .30. 8.4 ,

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7. 1.10 19. 5.2.2 31. 9.4
8. 2.8 20. 5.2.7 32. 12.1
9. 3.1.2 21. 5.2.8 33. 12.2
10. 3.1.4 22. 5.2.10 34. page V-3-17 I
11. 3.1.9 23. 5.2.12(1) 35. page V-3-11

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12. 4.3 24. 5.2.14 1

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7 Part C The questions in this part of the enclosure were not included in our

-letter dated December 12. 1968. However, the issues have been addressed in other related applications and have been discussed with your represen-tatives. Responses are needed to complete the'information supporting your application.

6.5 As a result of the Oyster Creek Nuclear Plant POL review, the auto-relief system was modified by the addition of an a.e.. interlock _in the actuation logie. This a.c. interlock would prevent actuation j of. the auto-relief system in the event a.c. Power is unavailable. j Please justify the absence of this a.c. interlock in your design.

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7.13 Please describe the differences, if any, between your design for /

the reactor protection system (including the initiating circuits for the engineered safety features) and that of the Cooper Nuclear Station (Docket No. 50-298) .

7.14 Please state whether the LPRM channels are a part of the reactor protection system. )

I 14.10 Please furnish additional discussion regarding the adequacy of the design bases and performance requirements of engineered safety features to ensure that the offsite doses in the event of an accident l will be within the guidelinea of 10 CFR Part 100. In particular, con-- )

sider the following h e engineered safety feature equipment considered in the analysse of doses to individuals offsite for the four design-basia seeidents described in Section XIV-3.0 (loss of coolant, refureling, control rod drop, and steam-line breek). Relate the design bases and par-formance requirements of such equipment to the assungtions used in the accident analyses.

The design bases for the control room ventilation system mentioned on page K-3-4 and for control room shielding given on page V-6-2.

De design-basis leak rate for the secondary containment and the design bases for the standby gas treatment system. Include in the discussion, the effects of exfiltration up to the design wind speed, the effect of the standby gas treatment fan capacity and the effect of time delays in starting this fan.

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