ML20207A118

Forwards Comments & Questions Re Emergency Planning Zone Evaluation to Be Forwarded to Applicant as Basis for Conversations Between Applicant,Nrc & BNL

ML20207A118
Person / Time
Site:	Seabrook
Issue date:	10/21/1986
From:	Berlinger C Office of Nuclear Reactor Regulation
To:	Nerses V Office of Nuclear Reactor Regulation
Shared Package
ML20205J677	List: ML20205J673 ML20205J692 ML20205J718 ML20205J743 ML20205J760 ML20205J769 ML20205J781 ML20205J794 ML20205J820 ML20205J852 ML20205J894 ML20205K095 ML20205K127 ML20205K168 ML20205K183 ML20207A118 ML20207A145
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FOIA-87-7 NUDOCS 8610300356
Download: ML20207A118 (5)

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FAI M.4 MEMORANDUM FOR: Victor Nerses Project Manager PWR Project Directorate #5, DPL-A FROM: Carl Berlinger, Chief Reactor Systems Branch Division of PWR Licensing-A

SUBJECT:

SEABROOK STATION RISK EVALUATION PERTINENT TO EMERGENCY PLANNING -

REFERENCES:

1. "Seabrook Station Risk Management and Emergency

,, Planning Study", Pickard, Lowe and Garrick, Inc.,

PLG-0432, December 1985.

2. "Seabrook Station Emergency Planning Sensitivity Study", Pickard, Lowe and Garrick, Inc., PLG-0465, April 1986.

Plant Name: Seabrook Station, Units 1 and 2 Docket Number: 50-443, 50-444 Responsible Branch: PWR Directorate #5 Project Manager: Victor Nerses Review Branch: Reactor Systems Rranch, DPL-A Review Status: Ongoing We have read the reference documents and have discussed their contents with personnel from Brookhaven National Laboratory (BNL), who are assisting the Staff in evaluation of the subject issue. We have agreed with BNL that we will follow certain potential containment bypass accidents while BNL will be responsible for other bypass conditions. This principally involves our taking responsibility for the steam generator as a bypass path resultino from core t

overfeatingphenomena,withBNLtakingresponsiM11tyforbypasspathswhich potentially initiate core melt accidents, such as LOCA outside containment.

We also agreed with BNL that we would pursue certain questions pertaining to the subject issue, and that they would purs'ue others.

The Enclosure to this memorandum represents our response to the split of work l between ourselves and BNL. We suggest this be forwarded to the Applicant with ,

the suggestion that it serve as the basis for conversations between the '

4 Applicant, ourselves, and perhaps BNL.

The ma,ior concern with steam generator behavior during a core melt accident is the rupture of multiple tubes in response to high Reactor Coolant System (RCS) temperatures which follow core uncovery. This accident sequence is of concern any time there is a core melt with the Reactor Coolant System at more than a few hundred psi pressure, with no water in the SG secondary side. These conditions le d to a potential for natural circulation transport phenomena to .

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,' OCT 2 f 1986 significantly heat the tubes prior to breach of the reactor vessel. The resulting loss of tube strength can lead to tube rupture. Reactor Coolant Pump operation, as outlined in many plant emergency procedures, almost assures this to be a concern. If tube rupture occurs, and any of the secondary side valves are open, the secondary side is breached outside containment, or the reactor coolant system pressure is above the SG relief valve setpoints, then containment is bypassed.

A number of questions need to be addressed on this issue in order to approach resolution with respect to the Seabrook investigation. Several of these are provided in the Enclosure. We anticipate further questions will arise as the '

investigation proceeds.

Original si.r:d by Carl Berlinger, Chief

- Reactor Systems Rranch Division of PWR Licensing - A Co,ntact: W. Lyon, x27940

Enclosure:

As stated

{

cc: C. Rossi R. Ballard Y. Benaroya l J. Milhoan

.' W. Minners C. Thomas M. Hodges S. Long G. Bagcht

8. Doolittle

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L key RSB:PWR-A RSR: WR-A b RSR:PWR-A on RLobel [CRerlinger WLy$o/86 10 10/ /86 10/g/86 Ao

( .

,.ro TECHNICAL EVALUATION OF THE EFZ SENSITIVITY STUDY FOR SEABROOK Draft Complete Contents Author at BNL

• Summary Pratt Nov. 7

1. Intr.o.duc t ion ,

Pratt Oct. 27

1.1 Background

1.1.1 Uniqueness 1.2 Scope and Focus of Review 1.3 Organization of Report *

2. System Evaluation 2.1 Interfacing System LOCA

, R.1.1 Operator Actions Luckas Oct. 27

1. 2.1.2 Break Location Berler  ??

2.1.3 Event Tree Quantificat)on Bozoki Oct. 27 2.2 Accidents During Shutdown and ' Refueling Conditions Chu Oct. 27

• 2.3 Induced Steam Generator Tube Rupture Lyon Oct. 30

• 2.4 Containment Isolation Failure Luckas Oct. 27 2.5 Completeness P. Davis Oct. 27

• 3. Evaluation of Containment Dehavior Hofmayor Oct. 31 3.1 Evaluation of Structural Strength 3.1.1 Capacity at General Yield 3.1.2 Capability of Penetrations 3.1.3 Behavior at Large Deformation 3.1.4 Summary of Structural Findings '

1. 3.2 External Events  ?? ,

??

1. • . 3.3 Treatment of Preexisting Leaks 77 77 1

4. Obntainment Event Tree 4.1 Sensitivity to Containment-Loads Chun Oct. 27

• 4.2 Sensitivity to Containment Performance Chun Oct. 31

1. 4.3 Sensitivity to External Events  ?? 77

5. Review of Source Terms khatib-Rahbar Oct. 27 ,

5.1 Fidelity to WASH-1400 Methodology 5.2 Credit for Scrubbing of Submerged Rele+ses p .rt)-97-7 d/yy l Sounrook contones 1 10/21/o6

f

o. bite Const quence Mredel 6.1 NUPEG-0396 Basis. Tingle Oct. 27 6.2 Consequence Modeling Tingle Oct. 27 6.2.1 Whole Body Dose vs. Distance .

6.2.2 Thyrold Dose vs Di' stance 6.2.3 Risk of Early Fatalities  ;

6.3 Time Before Release Comparisonc Tingle Oct. E7 w 6.4 Senuitivity Studies Tingle Nov. 7

• 6.5 Comparisons of Results Tingle Nov. 7 6.5.1 Retsults of Seabrook Study 6.5.H NUREU-0396 6.5.3 WNiH-- 1400 '

6.5.4 NRC Safety Goal

7. (Section 7. " Potential Improvements for Risk Reduction",

was removed by BNL.)

• indicates that date was changed or section way. reistated by NRC

1. indicates NRC wants information to be changed or added.

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I u En eh r or.l- c ri n t. o n t. < a 1 < ,/ 21/ n/,

+

f 10/22/06 AGENDA FOR BRIEFING ON '

REVIEW OF SEABROOK EPZ STUDY

1. BNL' Report

a. schedule

b. contents

2. Potential Weaknesses in PSNH Study i

a. Event V

1. check valve failure rates ,

2. scrubbing credit for submerged release

3. potential for becoming dominant contributor

b. S/G Tube Failure Due to Overheating i

1. company model -> will not occur

2. research work -> may be dominant risk j, contributor

3. Other Issues Significant to Review

a. BNL Assessment of Containment Challenge

b. PLG study on impact to risk from containment capability corresponding to 1% strain '

c. Dominant Dose Curve Release Catagory (S2W) Isotopic Ratio i 4. Caveats 1

a. Completeness l

b. Closeness to Criteria 4

1 fez /) ' $1~ 7 i L/fo i

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. Question 17 - Enclosure 3 UNITED 87ATES

1. k NUCLEAR REGULATORY COMMISSION

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wAsswoTom.o.c.seems Os" 2 3198t -

1 Do.ck. et Nos. 50-443/444 i

i Mr. Robert J. Harrison President & Chief Executive Officer Public Service Company of New Hampshire t Post Office Box 330 Manchester, New Hampshire 03105 i

Dear Mr. Harrison:

6

Subject:

Request for Additional Infonnation for Seabrook, Units 1 and Emergency Planning Sensitivity Study li The enclosed Request for Additional Information (RAI) is supp .

RAI dated October 8, 1986, except that question 20 is a restate question 20 posed in the earlier RAI.

. document the oral questions raised in our meet 1986. Please provide your responses promptly to facilitate our been included.

' review.

Questions or additional information regarding this matter sh to the Technical Project Manager for the review of the Seabrook Em i ,<

' PlanningSensitivityStudy,S.M.Long(301)492-8413.

Sincerely,

I s k if Steven M. Long, Project Manager PWR Project Directorate No. 5 i Division of PWR Licensing-A I

~

Enclosure:

As stated

' cc: See next page i

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Enclosure 1 l

REQUEST FOR ADDITIONAL INFORMATION SEABROOK STATION. UNITS 1 AND 2 4 DOCKET N05. 50-443 AND 50-444 l l EMERGENCY PLANNING SENSITIVITY STUDY f Restated Question:

1 l 20. Assess the impact on risk of assuming that the containment capability

corresponds to the pressure which produces 15 strain in the containment l wall.

l j Additional Questions: -

I l 29. The $2W release category isotopic distribution listed in the Table 4-3 in l PLG-0465 shows release fractions of cesius and tellurium that exceed the l release fraction for noble gases and greatly exceed the release fractions ll for elemental and organic iodine. Please justify the isotopic distri-ili*i. bution of the $2W release category consistent with WASH-1400 source tenn methodology, 1

i  !

! i l 30. The $7W release category isotopic distribution reflects a decontamination  !

j factor (DF) of 1000, for all isotopes except noble gases, because the ,

release point is submerged in the RHR vault. WASH-1400 source term i

, methodology credited BWR releases with a DF of 100 when they occurred l through a subcooled suppression pool, but set the DF to 1 when the pool was at saturation temperature.

l a. Discuss the degree of subcooling that would be expected in the RCS l water that pools in the RNR vault following blowdown through the RHR l system. i l

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b. Justify the use of a DF=1000 in light of the WASH-1400 methodology j and the degree of subcooling expected in the RHR vault water.

1 31. Provide a typical calculation to demonstrate that small diameter penetra-I tion sleeves do not punch through the containment wall under the worst l pressure condition assumed in the analysis.

32. In your prediction of large defomation behavior of the containment, full

! bond was assumed between reinforcement and concrete between two adjacent

) vertical cracks; assess the effect on containment behavior including pene-

] tration capability, if no bond f. tress is assumed between the reinforcing j' steel yield point and ultimate strength of steel. Based on our discussions

• in the meeting, it is our understanding that you will perfom this assess-

,f ment assuming no bond stress, f

33. Confim that a complete and independent check will be performed for the r

containr:ent strength calculations that served as the basis for the EPZ sensitivity study. i

34. Fully address the effect of uncertainty in the ultimate strength of Cadweld splices on the pressure capacity of the containment. As discussed ,

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jr in the meeting, your response should address potential, non-ductile '

l failure of the Cadweld splices.

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{' 35. Assess the response of the containment sump encapsulation vessel on the ,

{. containment integrity.

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j 36. Discuss the results of recent EPRI tests to address the potential for i

! strain concentration in the liner at crack locations. '

1 j 37. Demonstrate that your calculations fully account for the differences in

! stress-strain behavior between the reinforcing steel and the l h plate  :

I with regards to strain compatibility.

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38. Quantify the leak areas associated with other containment failure modes as discussed in Section 5 of Appendix H to the PLG report #PLG 0300. Also.

assess the impact on risk by assuming these failure modes to be type A l

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i j rather than type B failures including the effect of simultaneous occur-

] rence of various failure modes. ,

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l 39. Only selected penetrations were analyzed in the calculations; compile a

{ list of all containment penetrations, categorita according to behavior and j' demonstrate that each penetration is adequately covered by the analyses l l that have been performed.

4 j 40. What indications are available if RHR is lost during shutdown (e.g. '

spuriousclosureofsuctionvalve)?

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f 41. What indication is available for vessel level during shutdown and refuel-1 l* ing modes?

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l 42. Does loss of power to the pressure transmitter that provides input to the l autoclosure interlock for RHR suction valve cause the valves to close?

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} '43. Towhatlevel(s)istheRCSdrainedformaintenanceactivitieswhileshut-  !

down with fuel in the vessel? What level is necessary to maintain

! connection with the ultimate heat sink? l i

j, 44. Describe the availability of the $1 pumps while shut down. How difficult j, would it be to restore the SI function to respond to transients during  :

! shutdown and refueling conditions? Consider maintenance of the $! system

)- in your response.

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45. Provide the procedures for establishing cold overpressure protection when

shutting down.

l I 46. Is the primary system made water-solid during shutdown?

i I 47. Address the risk from creep failure of the steam generator (5/G) tubes due to exposure to high temperatures during core melt sequences in which the

reactor coolant system (RCS) remains at high pressure and the secondary

} sides of the 5/Gs are dry. Your discussion should reflect the recent experiments and modeling efforts that show 3-dimensional convective flows .

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_ .- ,- -= . . . - - = _ - . - - --- _ -

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,- 4 which transfer heat from the overheating core to other places within the RCS, particularly into the upper plenum and from the upper plenum along the hot legs into the S/Gs and through the U-tubes. Also include the influence of pressure driven flows resulting from reactor coolant pump (RCP) seal LOCAs, POPV/ safety valve actuations, " bumping" the RCPs, etc.

Localized heating effects due to redistribution of fission products in the t RCS should be included.

a. What is the total probability of occurrence for the high RCS pressure (

core melt sequence with dry S/Gs? ,

1,

b. What is the estimated conditional probability that the S/G tubes will fait due to overheating before the pressure is relieved by failure of the RCS elsewhere?

c. What is the effect of preexisting S/G tube leakage (within technical i

! specifications) on the heating rate and temperature required for t f failure of the leaking tube (s).  !

f i d. What release category would creep failures of the S/G tubes result  ;

in?

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lr 48. Most of the work pertinent to severe accidents has addressed plant behavior at full power, on the assumption that this represents the major i contribution to risk. Also, WASH-1400 assumed containment failure was i

probable following a core melt, making containment bypass sequences l I relatively less important. Therefore*

I a. Please address the possibility of accidents inside the containment l

! building while in Modes 2-6 (Startup, Hot Standby, Not Shutdown, Cold l Shutdown, and Refueling) insofar as these accidents could impact upon  !

risk. In particular, consider the effect of reduced safety equipment j

! availability and containment integrity requirements pomitted by l

! technical specifications while shutdown or refueling.

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b. Event V and steam generator tube rupture provide a direct path from the RCS to the environment during severe accidents. Please describe l the Seabrook work which identifies any other direct paths. l

c. Please provide further infonnation and/or specific references )

pertinent to release of radioactive material located outside of the l containment building (e.g. spent fuel pool, radweste systems) insofar as the magnitudes are large enough to impact upon the issue under consideration here.

49. The FSAR gives RHR relief valve flow rate as 900 gpm with a set pressure ,

of 450 psi. The flow rate does not agree with the value used in Reference 1, section 3, page 6. Please explain.

50. Please describe the mechanism for assuring that plant changes and new knowledge are promptly factored into the technical considerations which ,

fom a part of the foundation for staff consideration of a reduced emergency planning zone radius.

51. Reference 1, page 3-7, paragraph 5) references both high and low level

. sump alarms. What is a sump low level alam?

F ,

52. Page 3-7 contains a discussion of vault behavior in response to RHR system breaks. The emphasis is upon loss of equipment due to flooding. What consideration has been given to breaks which are small enough that the vault is not flooded, but there is a significant themal energy release that may impact equipment operation? Please include consideration that enough energy may be released to activate the fusible links in the ventila-tion system, thereby teminating ventilation and indirectly causing -

failure to pumps due to overheating of pump motors, and that this could l occur at a time earlier than might occur due to flooding. l l

53. Reference is made on page 3-7 to the RHR system crosstie line and RHR system response due to flow in this line as well as in the miniflow bypass lines. The conclusion is drawn that the RHR system pressure wjl1 tend to be unifom as a result. Are flow conditions such that this is realistic?

L-...-....- ......--. - . - - - . .

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j What is the impact of this assumption on conclusions pertinent to the i diseassion? ,

54. The authors conclude on page 3-9 that presence of water in the reactor cavity will decrease (significantly?) the revaporization of fission  ;

I products from RCS and perhaps RHR surfaces. We anticipate that a significant quantity of heat producing radioisotopes will remain in the wreckage of the reactor vessel, and this may be effective in heating what-ever gases or vapor are flowing toward the break. Has this been investi-f  ;

j gated?

l.

4 , 55. What is the justification for the statement on page 3-10 that the first 4

sign of trouble will be pressurizer low level or low pressure alams? We

. suspect a number of other indicators may be first, such as abnomal indications from the PRT or even a smoke alam.

j 56. There have been a number of indications (prior to and including page 3-11)  ;

that containment spray may be actuated due to RHR relief valve release  !

f into containment. What is the justification for this conclusion? Include the effect of containment heat sinks and containment cooler operation in I

the response.

57. The statement on page 3-11 that "As soon as the pumps begin to produce flow to the RCS valves in the miniflow lines close and all RHR pump flow is injected into the reactor vessel via the RHR cold leg injection lines" ,

i is not correct. The sensors are not located at the RCS to detect flow at f that location. Further, one is postulating a break in the RHR system, and

, a significant portion of the pump flow may never reach the RCS (as it l l statedinalaterparagraph). i i

i 58. The last paragraph on page 3-11 contains a number of timing of event  ;

statements, please provide justification of each. Plots of plant .I j behavior showing suitable parameters and indicating the event points are sufficient for most. Operator response infomation. in addition to RCS parameter infomation, is necessary to substantiate the statement that I RCPs will be tripped within about 21 seconds of break initiation.

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i 59. An item under consideration for advanced nuclear power plants is the

! ability to monitor pressure on the low pressure side of check valves.

This could provide early warning of check valve leaks and would provide

! monitoring capability to help assure check valves were operating properly.

The same monitoring capability with respect to RHR suction line valves

could identify if individual valves were mispositioned or malfunctioning.

! Would such a system for Seabrook be of significant benefit in reducing risk in a reduced size emergency planning zone? l

60. Please elaborate on the page 3-23 list of actions an operator can take to i mitigate the accident. This list appears to be short. Include identifi-I, cation of what has been incorporated into operator training and procedures

• l' at Seabrook.

i

61. What is the frequency of failures in the pipe tunnel that is mentioned on i page 3-23, and which led the authors to conclude they are very low?

62. Page 3-27 references situations where the combined sump pump capacity is j

sufficient to remove leaks and keep the vaults from flooding. In these l cases, the RHR, $1, and C5 pumps are assumed not to be impacted by flood-

! ing. Whatconsiderationwasgiventofailureofone(orboth)sumppumps?

i?

!I 63. What is the maximum flow rate that can be injected into the RCP pump lt seals? (Of potential interest since it may be an alternate path for

) injectionintotheRCS.)

64. Shutting an RHR system crosstie valve is identified on page 3-35 as an i action to help isolate a LOCA outside containment involving the RHR/51 i

systems. Has a careful evaluation of these systems been performed to l assess isolation strategy? If so, are procedures in place at Seabrook

! Station which reflect the work?

l 65. Relative water levels in the RHR vaults and the RCS are mentioned on pages

, 3 35 and 3-36. What are the water volumes in these regions as a function l l of elevation? (Of particular interest is the level at the top of the core i

! and at the elevation of the hot leg connections to the RHR.) l 4

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~~~ *~ L_2~~ : T _ r _- - _ - - _ - _ _ . . _ - . _--

66. What is the justification for the statement on page 3-36 that the water level in the vaults will be approximately the same as that in the RCS?

(We do not agree because of the potential that pressure in the. vaults and containment are not the same, and water temperature in the two locations may differ.)

67. Page 3-37 contains the wording "End state DLOC contains sequences in which the interfacing LOCA has been teminated, and the ECCS has been degraded (D) (RNR or Si pumps have failed)....The point estimate frecuency of DLCO is 4.0 x 10'I per year. The additional failures required to achieve core melt would lower this frequency by a least one order of magnitude." What ,

is the justification for this conclusion? (We have already lost a portion

, or all of the ability to inject water into the RCS via the usual paths.)

68. The bottom of page 3-37 contains a statement to the effect that failure of one charging pump will lead to core melt. Why is this the case? Our perception is that sufficient flow might be provided by alternate means to

. keep the core covered, such as use of the remaining two charging pumps, and perhaps the reactor makeup water pumps).

. 69. What is to be the status of the " temporary" 34.5 kV power lines which are

, identified on page 3-457

70. What is to be the status of the mobile power supplies which are identified on page 3-46?

71. What capability has been provided to connect external pumpt as identified in the second and third paragraphs of page 3-467 (This was briefly mentionedonpage3-48.) Use of a pump to simply inject water into containment via the sprays on a short term basis (no recirculation) does not appear to be identified. Has this been considered'

72. Page 3-46 identifies a number of possibilities for recovery of various safety functions. Are there specific plans? If so, please provide them.

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73. There have been several references to purchase of a mobile electric generator by pooled resources on the pages prior to page 3-49. What is the likelihood that such a generator would be needed by several plants at the same tii..a. ..d hence might not be available to Seabrook Station when needed? Similarly, where is the generator to be stored, and how is it to be transported to Seabrook? Include consideration of post seismic and post severe stom conditions in the response.

74. A tacit assumption appears to be incorporated into References 1 and 2 that ,

check valves are always closed. In reality, many check valves require a (substantial) reverse flow to force them to close, and they additionally 8

often require a significant reverse pressure to keep them closed. It this

. the case for any of the valves of interest here? If so, please discuss the implications. If not, what is the justification for the conclusion?

75. In the description of RHR pressure boundary failure modes it is stated that the maximum value of stresses due to pressurization to 2250 psia in the limiting RHR piping are approaching the yield stress and the stresses  !

in other metallic components are at a small fraction of their respective yield stresses. Describe the analyses conducted to support this conclusion and provide a sumary of the pertinent results. In addition, I clarify whether the pressure loading has been applied as a dynamic pulse I coupled with corrosion degradation effects (such as heat exchanger tube embrittlement). If these effects have been considered, describe the analyses and the dynamic loads. If not, provide the bases for not con-

• sidering these effects.

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