ML20215M395
| ML20215M395 | |
| Person / Time | |
|---|---|
| Site: | Seabrook  |
| Issue date: | 10/23/1986 |
| From: | Long S Office of Nuclear Reactor Regulation |
| To: | Harrison R PUBLIC SERVICE CO. OF NEW HAMPSHIRE |
| References | |
| NUDOCS 8610300213 | |
| Download: ML20215M395 (14) | |
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Docket Nos. 50-443/444 Mr. Robert J. Harrison President & Chief Executive Officer Public Service Company of New Hampshire Post Office Box 330 Manchester, New Hampshire 03105
Dear Mr. Harrison:
Subject:
Request for Additional Information for Seabrook, Units 1 and 2, Emergency Planning Sensitivity Study The enclosed Request for Additional Information (RAI) is supplemental to the RAI dated October 8,1986, except that question 20 is a restatement of the question 20 posed in the earlier RAI. Several of the questions in this RAI document the oral questions raised in our meetings on October 15 through 17, 1986. Additional questions pertaining to other areas of our review have also been included.
Please provide your responses promptly to facilitate our review.
Questions or additional information regarding this matter should be directed to the Technical Project Manager for the review of the Seabrook Emergency Planning Sensitivity Study, S. M. Long (301) 492-8413.
Sincerely, s
Steven M. Long, Project Manager PWR Project Directorate No. 5 Division of PWR Licensing-A i
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Enclosure:
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s g 3 OCT 1986 Docket Nos. 50-443/444 Mr. Robert J. Harrison President & Chief Executive Officer Public Service Company of New Fampshire Post Office Box 330 Manchester, New Hampshire 03105
Dear Mr. Harrison:
Sub. ject: Request for Additional Information for Seabrook, Units 1 and 2, Emergency Planning Sensitivity Study The enclosed Request for Additional Information (RAI) is supplemental to the RAI dated October 8,1986, except that question 20 is a restatement of the question 20 posed in the earlier RAI. Several of the questions in this RAI document the oral questions raised in our meetings on October 15 through 17, 1986. Additional questions pertaining to other areas of our review have also been included. Please provide your responses promptly to facilitate our review.
Questions or additional information regarding this matter should be directed to the Technical Pro. ject Manager for the review of the Seabrook Emergency Planning Sensitivity Study, S. M. l.ong (301) 492-8413.
Sincerely, Steven M. l.ono, Pro. ject Manager PWR Pro.iect Directorate No. 5 Division of PWR I.icensing-A i
Enclosure:
As stated cc: See next page Distribution
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Docket Nos. 50-443/444 Mr. Robert J. Harrison President & Chief Executive Officer Public Service Company of New Hampshire Post Office Box 330 Manchester, New Hampshire 03105
Dear Mr. Harrison:
Subject:
Request for Additional Information for Seabrook, Units 1 and 2, Emergency Planning Sensitivity Study The enclosed Request for Additional Information (RAI) is supplemental to the RAI dated October 8,1986, except that question 20 is a restatement of the question 20 posed in the earlier RAI. Several of the questions in this RAI document the oral questions raised in our meetings on October 15 through 17, 1986. Additional questions pertaining to other areas of our review have also been included. Please provide your responses promptly to facilitate our review.
Questions or additional information regarding this matter should be directed to the Technical Project Manager for the review of the Seabrook Emergency Planning Sensitivity Study, S. M. Long (301) 492-8413.
Sincerely.
f Steven M. Long, Project Manager PWR Project Directorate No. 5 Division of PWR Licensing-A
Enclosure:
As stated cc: See next page
Mr. Robert J. Parrison Public Service Company of New Hampshire Seabrook Nuclear Power Station cc:
Thomas Dignan, Esq.
E. Tupper Kinder, Esq.
John A. Ritscher, Esq.
G. Dana Bisbee, Esq.
Ropes and Gray Assistant Attorney General 225 Franklin Street Office of Attorney General Boston, Massachusetts 02110 208 State Posue Annex Concord, New Pampshire 03301 Mr. Bruce B. Beckley, Project Manager Public Service Company of New Hampshire Resident Inspector Post Office Box 330 Seabrook Nuclear Power Station Manchester, New Hampshire 03105 c/o US Nuclear Regulatory Commission Post Office Box 700 Dr. Mauray Tye, President Seabrook, New Hampshire 03874 Sun Valley Association 209 Summer Street Mr. John DeVincentis, Director Paverhill, Massachusetts 01839 Engineering and licensing Yankee Atomic Electric Company Robert A. Backus, Esq.
1671 Worchester Road O'Neil, Backus and Spielman Framingham, Massachusetts 01701 116 !.owell Street Manchester, New Hampshire 03105 Mr. A. M. Ebner, Project Manaaer United Engineers & Constructors William S. Jordan, III 30 South 17th Street Diane Curran Post Office Box 8223 Parmon, Weiss & Jordan Philadelphia, Pennsylvania 19101 20001 S Street, NW Suite 430 Washington, D.C.
20009 Mr. Philip Ahrens, Eso.
Assistant Attorney General State House, Station #6 Augusta, Maine 04333 Carol S. Sneider, Eso.
Office of the Assistant Attorney General Environmental Protection Division Mr. Warren Pall One Ashburton Place Public Service Company of Boston, Massachusetts 02108 New Hampshire Post Office Box 330 D. Pierre G. Cameron Jr., Eso.
Seabrook, New Hampshire 03874 General Counsel Public Service Company of New Pampshire Seacoast Anti-Pollution League Post Office Box 330 Ms. Jane Doughty Manchester, New Hampshire 03105 5 Market Street Portsmouth, New Hampshire 03801 Regional Administrator, Region i U.S. Nuclear Regulatory Commission Mr. Diana P. Randall 631 Park Avenue 70 Collins Street King of Prussia, Pennsylvania 19406 Seabrook, New Hampshire 03874 Richard Pampe. Esq.
New Hampshire Civil Defense Agency 107 Pleasant Street Concord, New Hampshire 03301
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Public Service Company of Seabrook Nuclear Power Station New Hampshire cc:
Mr. Calvin A. Canney, City Manager Mr. Alfred V. Sargent!
City Hall Chairman 126 Daniel Street Board of Selectmen Portsmouth, New Hampshire 03801 Town of Salisbury, MA 01950 Ms. I.etty Hett Senator Gordon J. Humphrey Town of Brentwood ATTN: Tom Burack RFD Dalton Road U.S. Senate Brentwood, New Hampshire 03833 Washington, D.C.
20510 Ms. Roberta C. Pevear Mr. Owen B. Durgin, Chairman Town of Hampton Falls, New Hampshire Durham Board of Selectmen Drinkwater Road Town of Durham Hampton Falls, New Hampshire 03844 Durham, New Hampshire 03824 Ms. Sandra Gavutis Charles Cross, Esq.
Town of Kensington, New Hampshire Shaines, Mardrigan and RDF 1 McEaschern East Kingston, New Hampshire 03827 25 Maplewood Avenue Post Office Box 366 Portsmouth, New Hampshire 03801 Chairman, Board of Selectmen RFD 2 South Hampton, New Hampshire 03827 Mr. Guy Chicheste*, Chaiman Rye Nuclear Intervention Mr. Angie Machiros, Chairman Committee Board of Selectmen c/o Rye Town Hall for the Town of Newbury 10 Central Road Newbury, Massachusetts 01950 Rye, New Hampshire 03870 Ms. Cashman, Chairman Jane Spector Board of Selectmen Federal Energy Regulatory Town of Amesbury Commission Town Hall 825 North Capital Street, NE Amesbury, Massachusetts 01913 Room 8105 Washington, D. C.
20426 Honorable Peter J. Matthews Mayor, City of Newburyport Mr. R. Sweeney Office of the Mayor New Hampshire Yankee Division City Hall Public Service of New Hampshire Newburyport, Massachusetts 01950 Company 7910 Woodmont Avenue Mr. Donald E. Chick, Town Manager Bethesda, Maryland 20814 Town of Exeter 10 Front Street Mr. William B. Derrickson Exeter, New Hampshire 03823 Senior Vice President Public Service Company of New Hampshire Post Office Box 700, Route 1 Seabrook, New Hampshire 03874
f Enclosure
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REQUEST FOR ADDITIONAL INFORMATION SEABROOK STATION, UNITS 1 AND 2 DOCKET NOS. 50-443 AND 50-444 EMERGENCY PLANNING SENSITIVITY STUDY Restated Question:
- 20. Assess the impact on risk of assuming that the containment capability corresponds to the pressure which produces 1% strain in the containment wall.
Additional Questions:
- 29. The S2W release category isotopic distribution listed in the Table 4-3 in PLG-0465 shows release fractions of cesium and tellurium that exceed the release fraction for noble gases and greatly exceed the release fractions for elemental and organic iodine.
Please justify the isotopic distri-bution of the S2W release category consistent with WASH-1400 source term methodology.
- 30. The S7W release category isotopic distribution reflects a decontamination factor (DF) of 1000, for all isotopes except noble gases, because the release point is submerged in the RHR vault. WASH-1400 source term methodology credited BWR releases with a DF of 100 when they occurred through a subcooled suppression pool, but set the DF to I when the pool was at saturation temperature.
a.
Discuss the degree of subcooling that would be expected in the RCS water that pools in the RHR vault following blowdown through the RHR system.
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- i b.
Justify the use of a DF=1000 in light of the WASH-1400 methodology and the degree of subcooling expected in the RHR vault water.
31.
Provide a typical calculation to demonstrate that small diameter penetra-tion sleeves do not punch through the containment wall under th;' worst pressure condition assumed in the analysis.
32.
In your prediction of large deformation 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 stress is assumed between the reinforcing steel yield point and ultimate strength of steel.
Based on our discussions in the meeting, it is our understanding that you will perform this assess-ment assuming no bond stress.
- 33. Confirm that a complete and independent check will be performed for the containment strength calculations that served as the basis for the EPZ sensitivity study.
34.
Fully address the effect of uncertainty in the ultimate strength of Cadweld splices on the pressure capacity of the containment. As discussed in the meeting, your response should address potential, non-ductile failure of the Cadweld splices, y
- 35. Assess the response of the containment sump encapsulaticn vessel on the' containment integrity.
- 36. Discuss the results of recent EPRI tests to address the potential for strain concentration in the liner at crack locations.
- 37. Demonstrate that your calculations fully account for the differences in stress-strain behavior between the reinforcing steel and the lower plate with regards to strain compatibility.
- 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
rather than type B failures including the effect of simultaneous occur-rence of various failure modes.
- 39. Only selected penetrations were analyzed in the calculations;' compile a list of all containment penetrations, categorize according to behavior and demonstrate that each penetration is adequately covered by the analyses that have been performed.
- 40. What indicatior.s are available if RHR is lost during thutdown (e.g.
spurious closure of suction valve)?
- 41. What indication is available for vessel level during shutdown and refuel-ing modes?
42.
Does loss of power to the pressure transmitter that provides input to the autoclosure interlock for RHR suction valve cause the valves to close?
- 43. To what level (s) is the RCS drained for maintenance activities while shut-down with fuel in the vessel? What level is necessary to maintain
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connection with the ultimate heat sink?
- 44. Describe the availability of the SI pumps while shut down. How difficult would it be to restore the SI function to respond to transients during shutdown and refueling conditions? Consider maintenance of the SI system in your response.
45.
Provide the procedures for establishing cold overpressure protection when shutting down.
46.
Is the primary system made water-solid during shutdown?
- 47. Address the risk from creep failure of the steam generator (S/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 seconoary sides of the S/Gs are dry. Your discussion should reflect the recent experiments and modeling efforts that show 3-dimensional convective flows
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. e 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, PORV/ safety valve actuations, " bumping" the RCPs, etc.
Localized heating effects due to redistribution of fission products in the RCS should be included, a.
What is the total probability of occurrence for the high RCS pressure core trelt sequence with dry S/Gs?
b.
What is the estimated conditional probability that the S/G tubes will fail 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 specifications) on the heating rate and temperature required for failure of the leaking tube (s).
d.
What release category would creep failures of the S/G tubes result in?
- 48. Most of the work pertinent to severe accidents has' addressed plant behavior at full power, on the assumption that this represents the major
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contribution to risk. Also, WASH-1400 assumed containment failure was probable following a core melt, making containment bypass sequences relatively less important. Therefore:
a.
Please address the possibility of accidents inside the containment building while in Modes 2-6 (Startup, Hot Standby, Hot Shutdown, Cold Shutdown, and Refueling) insofar as these accidents could impact upon risk.
In particular, consider the effect of reduced safety equipment availability and containment integrity requirements permitted by technical specifications while shutdown or refueling.
f b.
Event V and steam generator tube rupture provide a direct path from the RCS to the environment during severe accidents.
Please describe the Seabrook work which identifies any other direct paths.
c.
Please provide further information and/or specific references pertinent to release of radioactive material located outside of the containment building (e.g. spent fuel pool, radwaste 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 form 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 alarm?
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 thermal 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 terminating ventilation and indirectly causing failure to pumps due to overheating of pump motors, and that this could occur at a time earlier than might occur due to flooding.
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 will tend to be uniform as a result. Are flow conditions such that this is realistic?
f What is the impact of this assumption on conclusions pertinent to the discussion?
- 54. The authors conclude on page 3-9 that presence of water in the reactor cavity will decrease (significantly?) the revaporization of fission 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-gated?
What is the justification for the statement on page 3-10 that the first 55.
sign of trouble will be pressurizer low level or low pressure alarms? We suspect a number of other indicators may be first, such as abnormal indications from the PRT or even a smoke alarm.
- 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 into containment. What is the justification for this conclusion? Include the effect of containment heat sinks and containment cooler operation in 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"
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is not correct. The sensors are not located at the RCS to detect flow at 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 stated in a later paragraph).
- 58. The last paragraph on page 3-11 contains a number of timing of event statements.
Please provide justification of each.
Plots of plant' behavior showing suitable parameters and indicating the event points are sufficient for most. Operator response information, in addition to RCS parameter information, is necessary to substantiate the statement that RCPs will be tripped within about 21 seconds of break initiation.
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- 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.
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The same monitoring capability with respect to RHR suction line valves could identify if individual valves were mispositioned or malfunctinning.
Would such a system for Seabrook be of significant benefit in reducing risk in a reduced size emergency planning zone?
- 60. Please elaborate on the page 3-23 list of actions an operator can take to mitigate the accident. This list appears to be short.
Include identifi-cation of what has been incorporated into operator training and procedures at Seabrook.
- 61. What is the frequency of failures in the pipe tunnel that is mentioned on 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 sufficient to remove leaks and keep the vaults from flooding.
In these cases, the RHR, SI, and CS pumps are assumed not to be impacted by flood-ing. What consideration was given to failure of one (or both) sump pumps?
What is the maximum flow rate that can be injected into the RCP pump 63.
seals? (Of potential interest since it may be an alternate path for injection into the RCS.)
- 64. Shutting an RHR system crosstie valve is identified on page 3-35 as an action to help isolate a LOCA outside containment involving the RHR/SI systems. Has a careful evaluation of these systems been performed to assess isolation strategy? If so, are procedures in place at Seabrook Station which reflect the work?
What are the water volumes in these regions as a function of elevation? (0f particular interest is the level at the top of the core and at the elevation of the hot leg connections to the RHR.)
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- 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 terminated, and the ECCS has been degraded (D) (RHR or SI pumps have failed)....The point estimate frequency of DLC0 is 4.0 x 10-7 per year. The additional failures required to achieve core melt wo91d 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 l
on page 3-46?
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- 71. What capability has been provided to connect external pumps as identified in the second and third paragraphs of page 3-46? (This was briefly mentioned on page 3-48.) 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 time, and 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 storm 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 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, clarify whether the pressure loading has been applied as a dynamic pulse 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.
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sidering these effects.
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