ML20055J265
| ML20055J265 | |
| Person / Time | |
|---|---|
| Site: | Yankee Rowe |
| Issue date: | 07/27/1990 |
| From: | Sears P Office of Nuclear Reactor Regulation |
| To: | Jeffery Grant YANKEE ATOMIC ELECTRIC CO. |
| References | |
| NUDOCS 9008020003 | |
| Download: ML20055J265 (18) | |
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- Docket No.50-029 July 27, 1990 Ms. Jane M. Grant l
. Senior Engineer - License Renewal Yankee Atomic Electric Company i
580 Main Street Bolton, Massachusetts- 01740-1398 i
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Dear Ms.-Grant:
P
SUBJECT:
REQUEST FOR ADDITIONAL INFORMATION REGARDING YANKEE R0WE-SEVEREL i
ACCIDENT: CLOSURE-SUBMITTAL By--letter datea December-29, 1989, you submitted YAEC-1711, Yankee Nuclear.
Power Station Severe Accident.~ Closure Submittal. Based on the staff's: ongoing:
i review of the' Yankee.Rowe severe. accident closure submittal and its associated-documentation, the enclosed Aree lists of' questions for additional information are provided.
The questions are related to the individual plant MxaminationT(IPE), the L
containment performance improvement program,'and accident management. We' request that you provide written response to the questions in the enclosed List'1 within 1 month from'this letter.in order to expedite'our review process.
Oral responses are requested for the; questions in enclosed List 2 and List-3..
Meeting and/or telecons may be scheduled by contacting Pat Sears -(3012492-1436).
Questions on external event IPE analyses'will be prepared af ter _the completion' of the development of the IPE external event methodology,'which is targeted for September 1990.
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Sincerely, 1
Original Signed By:
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Patrick M. Sears, Project Manager Project Directorate 1-3 Division of Reactor Projects 1/11 Office of Nuclear Reactor Regulatio'n' i
Enclosures:
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.vocist No.50-029 I
!j Ms. Jane M. Grant Senior Engineer - License Renewal Yankee Atomic Electric Company 580 Main. Street Bolton, Massachusetts 01740-1398 Dear Ms. Grant *
SUBJECT:
REQUEST FOR ADDITIONAL'INFORMATION REGARDING YANKEE R0"
~RE ACCIDENT, CLOSURE SUBMITTAL i
By letter dated December 29, 1989, you submitted YAEC-1711, tankee Nuclear.
Power Station Severe Accident Closure Submittal'.
Based on the staff's ongoing review of the Yankee Rowe severe accidert closure submittal and its associated i
documentation, the enclosed three. list. cf questions-for additional information-are'provided.
1 f
The questions are related to the individual plant examination (IPE), the containment performance improvement (CPI) program, and accident management i
(AM). We request that you provide written response-to the first list of questions in one month in order to expedite our review process.
Oral responses are requested for the second and third list of questions.
Meeting and/or telecons may be scheduled by contacting Pat Sears (301-492-1436).
. Questions on external event IPE analyses will be prepared after the' completion of the develonment of the IPE external event methodology, which is. targeted' for September, 1990.
Sincerely, i
Patrick M. Sears,' Project Manager.
Project Directorate 1-3
/
Division of Reactor. Projects 1/11 /
Office of Nuclear Reactor Regula on
- l Encloiures: As stated cc w/ enclosures:
See next'page L:
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5 Mr. George Papanic, Jr.
Sr. Project Enfineer
- i Yankee Atomic Electric Company l
580 Main Street Bolton, Massachusetts 02111 Dr. Andrew C. Kadak, President
- l and Chief Operating Officer ankee Atomic Electric Company v
580 Main Street Bolton, Massachusetts 01740-1398 Thomas Dignan, Esquire 1
Ropes and Gray 225 Franklin Street i
Boston, Massachusetts 02110 l
Mr. T. K. Henderson Acting Plant Superintendent Yankee Atomic Electric Company.
l Star Route 1
Rowe, Massachusetts 01367' i
Resident inspector Yankee Nuclear Power Station U.S. Nuclear Regulatory Commission-Post Office Box 28 Monroe Bridge, Massachusetts 01350
- Regional Administrator, Region 1 U.S. Nuclear Regulatory Comission
- 475 Allendale Road-1 1
King of Prussia,-Pennsylvania 19406 Robert M. Hallisey, Director Radiation Control Program Massachusetts Ocpartment of Public Health 150 Tremont Street, 7th Floor Bosten, Massachusetts 02111-Mr. George Sterzinger Comissioner Vernant Department of Public Service 120 State Streat, 3rd Floor t
Montpelier, Vermont -05602 1
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s ENCLOSURE REQUEST FOR ADDITIONAL INFORMATION P'EGARDING YANKEE R0WE-SEVERE ACCIDENT CLOSURE SUBMITTAL LIST 1 i
1.
Page 10-82 of Probabilistic Safety Study-(PSS).= You-state that-except for the TPLB sequence, there is only a 3% chance that the core; debris.will not be coolah.le in-vessel. Concisely discuss how you have made this determination.
Have procedures required to achieve this objective been prepared and implemented?
Page 10-82 of PSS: You state that because of hot leg or cold leg injection, yom can circumvent any blockages that may exist on the bottom of the reactor vessel.
Please explain this further.
Frr-instance, how are you assured that a crust will not form'thus keepirr W cooling water from mixing with the debris?
2.
Page 117'of. Closure Submittal: Discuss how core catcher ieatures, such as the ones shown in Figure 5-2, " General Arrangement Reactor Cavity Drain Line Plugs and Stand Pipes" will function to mitigate the effect_on containment of vessel melt through.
Previde information on the concrete-composition-(limestone vs. basalt).and the thickness of the shield tank walls.
Describe the location and purpose of the. top supporting ring of the sH eld tank. Has concrete been removed from the' area under the vessel, as the figure seems to indicate? Also, discuss the results of the-water injection to the cavity investigation-mentioned on page 117.
Page 124 of Closure Submittal:
Fer recirculation purposes, where is the location of the containment sump? Explain the functioning of the plugs (and-their material of construction) under the vessel support area;;also, describe the function of the neutron shield tank in' severe accidents..
3.
Page 10-4 of PSS: You acknowledge that "there is a high likelihood of the operatino staff of YNPS taking action-to align and actuate manual systems." Are there approved procedures in. place forntaking these actions?
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There. is an 88% probability that the containment will not Tail (thus,.
L Release Cat. #5).
For these cases, significant hydrogen can be produced before an in-vessel,-coolable, low-temperature, debris bed can be formed.
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Discuss how hydrogen can be produced and what the effects are of subsequent burns (if combustion cor.ditions can be achieved) on containment 1
integrity. Discuss how radioactive shine effects from the containment' impacts cruciri activities.
Page 10 P.t.
i PSS:
Concerning. hydrogen ignition, discuss what. ignites the hydrogeri : urns Sr both the ac-available and the ac-not available cases.
Discuss the realts of any analyses that you have on the effect'of hydrogen burns (that a not fai1~ the containment by overpressure) on the integrity of the containment, particularly as the temperatures might affect the leak tightness of the penetrations.
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Page 10-83 of PSS: You state that catastrophic (vessel) failure for the high pressure case has been deemed highly improbable and thus "not treated any further."
Is that your present position and, if so, is it bcsed only i
on Ref.10-127 6.
Page 10-86 of PSS: Discuss the temperature effects that were assumed for the containment failure analysis.
Page 4-82 of PSS: What are the environmental conditions assumed for the isolation valves curing a severe accident?
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7.
Page 10-81 of PSS: You state that conditions in the primary system negate the possibility of a steam explosion. Explain why.
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8.
Page 10-88 of PSS: You appear to consider containment isolation failure 1
about half-way through your event trees.
Normally it is considered much earlier in the event tree. The most important time to consider this so-called " beta" failure mode is for cases that would otherwise not fail, that is those that result in Release Cat. #5 from Event N ee Figure 10-6.
Why did you exclude containment isolation failure from the most benign and most probable release. category? Also, noting the Event Tree in Figure 10-9, why do you assign a release ~ category of #5 to a situation where you have a failed containment, only passive containment cooling, and a 3 MW heat source in containment? How does 98% of the noble gases remain in the containment?
9.
Page 10-88 of PSS:
In Section 10.8.14, you determined that containment failure is unlikely because only 7% of the concrete thickness has been-lost.
In Section 10.8.15, you determine that 100% of the concrete thickness has been lost but say nothing about the accompanying possibility that the containment will fail.
Discuss the apparent inconsistency.
- 10. When the containment fails by " major structural collapse," what is the final location of the core debris? Why is the duration of the release only 30 minutes?
11.
Page 131 of Closure Submittal: What effect does the early melting of.the control rods / safety rods have on recriticality? On core melt progression?
- 12. Provide a thorough discussion to justify why the Yankee IPE was performed using a DRA that is out-of-date.
Discuss how Yankee incorporated the most current deign and operations-information into the IPE and whether the results and conclusions presented reflect the current plant.
Certify that the IPE represents the as-built, a;-operated plant.
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Provide the basis for the conclusion that the IPE as submitted accounts for the effect of all modifications made to the plant subsequent to the-freezing of the design'(19811 to perform the Yankes Rowe PSS. Your IPE submittal (Section 3.3.4) dest. J several plant moc!fications performed since publication of the PSS in 1983.
Provide a concise list of-all; safety related plant modifications (both those motivated by the PSS and those made for other reasons) since 1981 (when the PSS design was frozen) and describe the potential downside, if any, of these modifications.
- 13. Discuss how an indepe dent in-house review was conducted to ensure the accuracy of the !?E documentation package and to validate both the IPE process and its results. Provide, as a minimum, a description of the internal review performed, the results.of the review team's evaluation, and a list of the review team members.
Describe the walkthrough/walkdown activities (e.g., initial walkthrough for plant familiarization; special.ones to verify logic l trees, dependencies, or aspects of systets interactions; to examine spatial interactions such as internal flooding)-including scope and team makeup.
Describe Yankee Nuclear's involvement in the plant walkthroughs/walkdowns.
- 14. Provide a thorough discussion of the evaluation.of'the decay heat removal function to address resolution of the USI'A-45 " Decay Heat Removal 3
- Requirements." The discussion should identify and quantify the-contributions of USI A-45 to core damage frequency or unusually poor containment performance.
- 15. The Safe Shutdown System (SSS) appears to enhance Yankee's -safety capability significantly, especially with regard to external events.
Provide the probabilistic assessment of the availability of the SSS and.
treatment for human recovery actions at the SSS for'the leading sequences requiring use of-the facility.
16.
Define core damage as used in the Yankee IPE.
Provide a description of how vulnerabilities were defined and identified.
2 Discuss the fundamental causes of any vulnerabilities identified.
List 4
the core damage and containment failure sequences that were selected by r
the screening criteria (Appendix 2) of Generic Letter 88-20. Provide a concise discussion of the level at which the criteria-were applied'(e.g.,
vj systemortrain).
- 17. The IPE Generic Letter requested licensees to discuss unique plant
- features thht contribute significantly to improved or reduced core damage frequency or good containment performance. Although your IPE submittal did not address this area explicitly, we believe that there are several-plant features (e.g., the Safe Shutdown System and the passive Vapor Containment) that should be highlighted.
Provide a concise discussion of those unique plant features that Yankee Rowe believes provide a iium
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. l substantial safet,, benefit. Describe those. features that require special attention by operating personnel.
- 18. Provide train level dependency tables / matrices for dependencies between front-line and support systems as well as for dependencies among support systems.
(Note:
this is not a single-failure analysis.) Particular attention should be paid to de power, component cooling, service water,
. room cooling, control air, and pump lubrication.
Identify where special dependencies were accounted for'in the IPE internal events evaluation. What sequences. leading to core damage were affected.by spatial-dependencies?
- 19. Provide the appropriate minimum success criteria'for event trees / front-line systems used in the IPE.
Indicate the basis for these criteria and the degree of conservatism (or whether best-estimate or optimistic) used.
Provide-the success criteria for initiating events developed in the master logic diagram. A statement for example-that "high pressure safety injection (is required)" does not indicate if one, two, or three charging pumps are required or if only two safety injection pumps are required or perhaps a combination of both is satisfactory, l
- 20. Provide'a concise description of how and why the' component cooling water, l
service water, and control air system failures were lumped.into the' Plant Trip initiating event tree.
- 21. Provide a concise discussion of how the initiating events frequency for non-isolable LOCAs was determined for the IPE.
Provide a concise discussion of how intersystem LOCAs were evaluated under the IPE for the shutdown cooling system.
- 22. The internal flooding analysis described in the Yankee Rowe IPE submittal-L did not clearly explain how the intent of-the IPE evaluation on internal flooding was met by the work previously performed for Yankee Rowe.
Provide a concise discussion of how internal-flooding was evaluated in a systematic manner.(such as by a PRA) and whether the following were
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l-considered: overfilling water tanks, hose ruptures, and pump seal leaks.
Provide references to the assumptions and methods used in the internal flooding analysis.
Identify critical areas where flooding may be a concern. Provide an estimated frequency of flooding initiation as well as estimated flow rates.
Identify recovery actions credited in the flooding analysis.
Identify resulting damage to important. equipment from various floods. Quantify and report the flooding sequences leading to core i
damage.
List important flooding sources.
- 23. Discuss the impact of loss of service water on plant systems, and estimate its contribution to core damage frequency.
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5-z 24 Discuss the need for feed and bleed, the probability of success, and the.
impact on are damage frequency from loss of this function.
- 25. Discuss the containment nodalization process used to effectively treat the.
accident progression back-end analysis in the Yankee IPE.
E6. Discuss how hydrogen pocketing was treated in the Yankee IPE..
Discuss how the Yankee IPE investigated the potential of damage to containment and
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important equipment as a result of local hydrogen detonation / deflagration.
i 27.
Provide a concise description of the additional systems being considered to retain the reactor core within the vessel,=as mentioned in'the CPI portion of the submittal presented at the May 3, 1990, meeting.
28.
Provide the remaining summary sheets for major systenis similar to the ones provided for main feedwater and recirculation. The major plant systems identified are: HPSI, LPSI, accumulator, chemical shutdown, emergency 1
feedwater, reactor protection system, and the containment isolation l
system. As depicted in the May 3, 1990, meeting handouts, the system sunenary sheets include: mission times, success criteria, failure probability, and major cut set contributors.
- 29. No discussion was found concerning the phenoniena of direct containment heating-resulting from high pressure melt ejection. Discuss the extent to which DCH was considered as part of high pressure reactor vessel failure i
scenarios.
- 30. Steam generator tube rupture has emerged as a major contributor to bypass leakage.
It is listed as one of the initiating events that were examined
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in the Yankee PSS. No mention was fcund, however, concerning the possibility of induced steam generator tube: rupture. Discuss.the extent to which steam generator dryout induced SGTR was considered.
- 31. Discuss the factors which influenced the-final selection of the Yankee plant damage state bins.
In other words, discuss how you assured yourselves of their adequacy in capturing all important plant conditions.
4 Yankee PSS Figures 10.6 through 10.9 contain the end states of the containment event tree. List the final conditions of temperature and pressure in the containment for those many end states labeled "OK," as well as.the conditions that resulted.in release categories 1 through 5.
32.
Concerning Yankee containment failure modes, explain the difference between reactor cavity failure upon debris solidification, reactor cavity failure due to melt-through, and'early failure of the reactor cavity. Do these failures imply concurrent containment failure or significant containment challenge?
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.. The' Yankee PSS contains six containment failure mode;, none of'which appear to address vulnerability of penetrations to thermal attack.
Show how this~ concern was resolved in arriving at the final modes of
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containment failure.
Were containment temperaturos less, in all accident scenarios than the high temperature envelopes which the penetrations were tested -to, as described in Yankee PSS Appendix E?
- 33. Discuss, in a paragraph or two, the structural analysis, based on a failure criterion of 0.9 yield stress, used to estimate the containment i
overpressure capability. Was it an in-house nalysis? Finite e1ement
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analysis? When was it performed? Have containment modifications been made.since the time of the original analysis, and if so, are the results unchanged?
- 34. Discuss the containment separate effects studies conducted from'1987 to the present. What was studied.and wny? What was learned?'
- 35. What factors influenced the decision to use plant-specific daca, generic-data, or a combination of both in various applications,in PSS?
How would the conclusions of the IPE hange if a more up-to-date data base instead of the NPRDS reliabi h ty data: base were used to provide generic data? What is your basis for this conclusion?
What were the sources of-generic error data from which bounding values were derived for estimating human errors? Why.were these sources selected?
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36.
In light of our understanding today of human reliability, manufacturing.
defects, and maintenance errors, provide a concise discussion of.why common mode / common cause failures were found.to be negligible contributors to unavilability/ failure in each of the 40 major system fault trees.
List the component groups subjected to CCF analysis.
Provide a concise discussion of the sources of CCF rates _used in the IPE.
When was the human reliability analysis.first conducted? When was it-updated? Was it requantified to account for the change to symptom based J
. procedures?- To what extent does it consider common cause? Does it take advantage of recent work on common cause?
- 37. What percent of the: core damage frequency was due to human error? What inferences (insights) are drawn regarding the contribution of human error-on overall plant risk?
Identify those sequences that, but for low-assumed human error rates in recovery actions, would have been above the screening criteria of Appendix 2 of Generic Letter 88-20.
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- 38. List the most dominant human recovery actions identified in the Yankee IPE, along with-the task analysis performed for each.
What type of human systems analysis was performed to support plant model development and to identify pertinent human task actions for inclusion in the event and fault trees.
What types of task actions.(both cognitive and physical) were analyzed as part of each accident sequence? How were -they chosen?
- 39. Provide a concise discussion to justify the operator actions without procedures for which the IPE takes credit. Quantify their contributians to the likelihood of core damage or containment failure.
Whatperson-centered (e.g.,) experience, fatigue, stress), task-centered (e.g..
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team support,. organizational support) performance shaping factors (PSFs),
were scaled for human task actions included as precursors, initiators, or meaiators in the event'and fault trees? How were they chosen? What tethods were used to scale each PSF? Why were those methods chosen?
What quantification methods (e.g., THERP, HCR, SLIM-MAUD) were used to l
l estimate human errors on task actions selected for analysis? Why were these quantification methods chosen?
- 41. Was a sensitivity analysis of human error performed? What characterization or behavioral model of plant personnel was used to identify multiples and dividends of the base or point estimates of human error for the sensitivity analysis?
- 42. To what extent were results documented to allow for auditing and/or l
replicating, or to allow for combining with data from other PRAs?
How was the completeness of the set of human faults verified?
- 43. Does the PSS consider maintenance induced events?
44._ Provide written assurance that the procedures and operator actions for which the IPE takes credit ara.in place at Yankee Rowe and that the operators have received training on these proceaures.
45.
Provide a lir.' of the equipment for which plant-specific data were used in c
the PSS and provide the plant-specific data failure or initiating event rates particular to the equipment.
- 46. Provide a concise description of how the plant-specific data were combined with generic data, particularly when there were significant differences between the rates or.when there was a statistically significant amount of plant-specific data for a particular component or system.
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REQUEST FOR ADDITIONAL INFORMATION REGARDING YANKEE ROWE SEVERE ACCIDENT CLOSURE. SUBMITTAL LIST!2 1.
Describe in & concise manner how your JPE' evaluation considered ventilation / room cooling failures. What were the conclusions in the IPE of the significance of these initiators?
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The PSS states in Section 5.3.1.1 that the Increase in Steam Flow initiating event only considers increased flow from a single steam generator. Given events.at the other-PWRs when more than one steam generator relief valve has failed simultaneously, what is the basis for 1
only evaluating the failure for one steam generator?
3.
The Rod Ejection initiating event was subsumed into the Intermediate LOCA initiating event. How was the increase in-reactivity from the rod i
y ejection factored into the Intermediate LOCA initiating event analysis?
blowing down all four steam generators (thereby drying them out) and then having cold ' emergency boiler feedwater pumped into the hot, dry steam generators and possibly causing SGTR7 5.
Provide a concise-discussion of why it was determined that there was no need to generate fault trees for the component cooling water system, service water system, and the control air system.
6.
In which sequences leading to core damage were environmental effects an important contributor?
7.
Concisely describe how the IPE evaluation has addressed the survivability of equipment that may be required to perform a mitigative function in a post-core melt environment.
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Concisely describe how check valve-failures were modeled in the IPE, including the assumed test interval.
9.
In light of the long operational history of Yankee Rowe with its attendant abundant amount of plant-specific data, what is the justification for using generic data failure rates for equipment that has a significant data.
history (e.g., ECCS and shutdown cooling pump)?
- 10. Provide a concise discussion of why plant-specific data were not used in a i
Bayesian update for the following equipment:
SG safety valves fail to open
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SG safety valves fail to close CSV/LSV fails to open CSV/LSV fails to close Atmospheric steam dump valves
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! i 11.. Provide a concise description of the period of operation from which plant-specific data was used.
If the period differed significantly by component, provide the range of dates and indicate the period of plant experience under which most of the equipment falls.
- 12. ~ Table 7-2 of the PSS describes the generic failure rates used in the PSS.
Provide a listing of the data sources referenced in the table under the heading " Data Source".
13.
Provide a concise discussion of how the EBF steam-driven pump is controlled following a station blackout. Discuss how this control is maintained (including the instrumentation relied upon) with and without dc power. - What battery depletion time was considered for station blackout sequences? What was the basis for this time?
14, Provide a concise description of how the IPE has included the possibility of the closure of the main coolant loop isolation valves which would prevent safety injection water from reaching the reactor vessel.
- 15. Discuss how the.IPE has taken credit for the use of the main coolant loop isolation valves to mitigate either a LOCA.or steam generator tube rupture. Has it taken credit for its use in any sequence? Has the IPE considered the inappropriate istiation of a main coolant loop?
- 16. Discuss the position of isolation and check the-valves in the shutdown cooling system in relation to containment boundary, surveillance methods and frequency, ability to isolate breaks, administrative controls on the shutdown cooling system MOVs, the design rating of-the piping, and a list of vital equipment subject to failure due to flooding from a break-in the shutdown cooling system.
- 17. One of the main goals of the Individual Plant Examination for a nuclear power plant is to identify potential vulnerabilities. ' Plants with unique design features may also have unique vulnerabilities. The following unique features to Yankee Rowe are potentially safety-significant and deserve additional at*cntion:
..The low pressure surge tank (LPST) acts in a similar manner to the pressure relief tank at other PWRs, but the LPST is outside of containment at Yankee Rowe.
Provide a concise description of how the IPE evaluates a stuck open pressurizer safety. valve or PORV/ block valve that will blowdown to the low pressure surge tank. How does the IPE take into account the design pressure of the tank and tank / rupture disc / piping failures that might cause an intersystem LOCA outside of containment? Provide a concise description of how an ATWS at Yankee Rowe would affect the low pressure surge tank. What events does the IPE postulate that could fail the surge tank or-its piping such that core damage and containment bypass would result?
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....The Yankee Rowe plant has loop isolation valves in the main coolant lines.
Provide a concise description of how the IPE has accounted for the status of the RCS loop isolation valves during power operation and isolation valve testing. How has the IPE considered isolation valve testing.
How has the IPE considered any administrative controls or Technical Specifications that affect these valves? Concisely describe how problems with valve disc failure on similar valves has been incorporated into the IPE.
Concisely describe how the E0Ps and/or the IPE take credit for use of these valves.
..At most plants the emergency feedwater pumps receive automatic start signals, but not so at Yankee Rowe. Concisely describe "a the IPE has considered this situation. Since closure of the mainsteam non-t turn valves will isolate the normal steam supply to the turbine-driven EBF pump, what signals will automatically isolate the steam supplies to this pump? What assurances are there that the MSIV bypass valves are able to provide motive steam to the EBF pump in the event the non-return valves close? Has the IPE modeled the power needs of the bypass valves, considered from where the valves must be controlled, and considered how valves in the turbine-driven EBF system fail on loss of motive power? How has the IPE considered the possibility that the turbine-driven pump may trip on high exhaust pressure?
..At most other plants it is not possible to feed _the secondary side of steam generators from primary systems pumps, but this is the case at Yankee Rowe.
How does the IPE take into account the time it takes to manually initiate the introduction of flow to the steam generators using the charging pumps?.
Similarly for the high pressure safety injection-and low pressure safety-injection and low pressure safety injection pumps as a backup method of supplying feedwater to the steam generators?- How does the IPE take into account the frequency with which the locked closed valves in these lines' are exercised / tested?
..Most plants do not have a dedicated, bunkered safe shutdown system that can supply makeup to the primary system and remove decay heat through the steam generators, but Yankee Rowe has this capability.
How has the IPE taken into account the contre 11ed capacity (by Technical Specification or other controls) of the fire water storage tank as a source of water to remove decay heat via the safe shutdown system? How has the IPE taken _into account the frequency with which the safe shutdown is tested? How has the IPE taken into account the safe shutdown system in general?
..At most plants, the emergency feedwater system is automatically loaded onto an emergency bus un loss of normal power, but at Yankee Rowe this action has to be performed manually. How does the IPE take into account the amount of time it takes an operator to go from the cantrol room to the area where the EBF is locally actuated? How does the IPE take into account how long it takes an operator (does it take more than one?) to align the EBF to a 1E bus? Is this done locally, in the control room, or by the operator sent to actuate the EEF I
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locally?. Huw does the IPE take into account the minimum number of operators required to be in the control room or in the plant? How does the IPE take into account the margin (time) to steam generator dryout between loss of all normal feedwater on loss of offsite power and restoration of an ac motor-driven feedwater supply?
..We cannot find the following information on Yankee Rowe: the capacity of the motor-driven EBF pumps, the number of steam generators that must be operable to adequately remove decay heat with EBF, and how long minimum levels required by Technical Specifications)(based on tank capacity and the plant can remain at hot standby and then cool down to cold shut down. P? ease supply this information.
..Since the mo_ tor-driven E8F pumps are not automatically loaded on the emergency diesel generators, we assume that there is.a capacity concern with these DGs. How has the IPE taken into account the limitations on equipment that can be simultaneously loaded on the diesel generatcrs-given a loss of offsite power and either a small break LOCA or a transient? How does the IPE take into account and do the E0Ps indicate what, if any, combinations of equipment should not be simultaneously loaded on/ started from the diesel generators?
..Most plants do not have pressure relief valves at the suction to the steam generators in the feedwater lines, but Yankee Rowe does.
How does the IPE take into account these valves and when they may lift, for example, due to pump discharge pressure (e.g., charging pumps feeding the secondary side)?
..Most PWRs have niultiple. accumulators to meet ECCS acceptance criteria and Appendix K criteria, but Yankee Rowe has only one accumulator that is normally kept depressurized. Does your best-estimate ECCS calculation indicate that you can get by on a large break LOCA (e.g., less than 2200 degrees F PCT) without-your single accumulator?
Is it needed for a medium break? How did the IPE take credit for the accumulator.and evaluate its pressurization feature on a I
safety injection actuation signal?
..Few PWRs today have all manual realignment for going from the infection to the recirculation mode of ECCS. How did the IPE take into account the manual steps (including where they must be performed and the time required) required to realign the safety injection system to recirculation? At most PWRs there are redundant ECCS sumps inside of containment in case of debris blockage of one sump. Are there redundant recirculation sumps at Yankee Rowe? Since switchover to recirculation is a manual action, how does the operator know it is time to begin recirculation and how much time does he have to complete his tas k? -
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REQUEST FOR ADDITIONAL INFORMATION REGARDING YANKEE R0WE SEVERE ACCIDENT CLOSURE SUBMITTAL LIST 3 Procedures 1.
Provide more.information regarding revisions to emergency operating procedures (E0Ps) on inadequate core cooling (ICC) which are said to be under evaluation on p.136. What will the decision criteria be for implementing any modifications.
1 2.
Westinghouse emergency response guidelines. (ERGS) are intended for conditions up to inadequate core cooling. Describe how YNPS E0Ps extend beyond ICC;. specifically:
. What depths are backup systems and alternate system configurations called out in E0Ps or other guidance?
. Are strategies for dealing with accidents beyond the DBA provided within or external to the E0Ps?
Are the. interfaces /transitiont-between E0Ps'and A/M procedures clearly identified?
What is the form and content of accident management (A/M) l
- procedures and guidance provided to 'the control room operators?.
technical support staff?
Guidance and Computational Aids 1.
The assessment of core damage based on the. post-accident sampling system (p. 142).is time consuming and could significantly lag.a transient.
Is additional damage, (b) guidance provided at YNPS for (a) projecting. time to core-assessing extent and type of core damage, including.the state and location of the core, (c) cooling the coro or-debris.following core damage or reactor vessel failure?
l 2.
Describe any structured guidance or computational aids provided'to the i
i operators or technical support staff at YNPS for estimating or-projecting:
leak rates and paths from~ containment and reactor coolant system 1
(RCS)
L reactor power level (via indirect measurements) injection flow rates into-the RCS, steam. generators, or containment for alternate system configurations state of core / debris, including extent of oxidation, location, coolability remaining time to key ' events, such as battery. depletion, core
'I uncovery, core degradation, reactor vessel failure, containment' failure pressure containment and reactor building atmosphere flammability and-detonability-anticipated pressure rise from hydrogen combustion or reactor vessel failure
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Training.
1.
Describe how the personnel and resources of the Nuclear Services Division (NSD) are typically involved in emergency response. Specifically, Are NSD personnel formally a part of the emergency response organization (ERO)?
How many-people and disciplines are included?
-i What positions do'they-fill in the on-site and corporate ER0s?
How frequently do these personnel participate in emergency preparedness drill and exercises?
What plant parameters are provided in the corporate office?
2.
Provide a description of the training programs / courses offered to YNPS personnel on severe accidents and accident management.
For each course include a description of training objectives, course outline, technical content (such as severe accident phenomenon, PRA/IPE insights, and A/M strategies), number of course hours allocated to each training' topic, and whether course has been reviewed by INP0 as part of training accreditation process.
Identify (preferabl 3.
required (by YAEC) y via a training matrix) which of the above_ courses are for each position in the on-site and corporate emergency response organization, and the frequency of this training.-
Consider separately the training requirements for control room shift personnel including shift supervisors, reactor operators, and shift-technical advisors; technical support staff including mechanical, t
electrical,.and ther.nal hydraulic specialists; directors'and/or coordinators of th plant emergency organization; and other utility j~
managers or decision makers who may play important roles-in emergency response..
l L-4.
Discuss how the training needs for ERO personnel have been established, and the extent to which a systems approach to training (including job and tasksanalysis)hasbeenused.
5.
Describe the process used by YAEC-in the last 3 years to develop scenarios l
for use in emergency preparedness exercises and drills.
In particular:
l l-Were scenarios developed in-house or by contractors?
To what extent _were key participants in the IPE program (i.e.,
principal authors and contributors) involved in scenario development?
To what extent were IPE results and insights reflected in the scenarics?
Were alternate success paths explicitly identified as part of.
scenario development? Was this information disseminated in any; way within the YAEC organization?
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Describe how emergency preparedness exercises and drills are used at YNPS to enhance A/M capabilities.
In particular, please discuss:
The extent to which preventive and mitigative measures-identified by operators and technical support staff are simulated or played out during drills and exercises.
1 The process by which preventive and mitigative measures identified during drills and exercised are reviewed as part of the exercise' critique-or followup activity, and the need for l
relaied plant / procedure / training enhancements is assessed Examples of plant / procedure / training enhancements implemented at
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'YNPS as a result of feedback from exercises and drills.
The use of equipment mockups to add realism.
7.
Describe any YAEC initiatives in the use of emergency exercises for A/M training, for. example:
Use of e.ultiple success paths in exercise scenarios to allow better simulation of recovery actions Conduct of success - oriented drills such as.INP0 casualty:
l control drills Use of the plant simulator to drive exercises and drills 8.
Describe any YAEC initiatives or plans for the use of plant simulators, nuclear plant analyzers, or tabletop-exercises.to enhance A/M capabilities, for example:
Extension of simulators beyond inadequate core cooling Use of plant simulator to drive exercises and drills Use of IPE results in the development of simulator-based teaining programs Development of training programs to address decisionmaking during. accidents A/M Plan 1.
Please provide a copy of the formal Yankee accident management plan or policy statement referred to on pages 143-14t.,
Discuss whether YAEC considers this a binding commitment.
2.
Based on the submittal (e.g., first paragraph in Section 6.5) it appears that the five steps in Yankee's A/M plan are only in.the planning stage.
-Please discuss the status of each step of the evaluation and provide an approximate schedule for any planned future activities.
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DISTRIBUTION:
Deeke%:fMe Q439;yp NRC PDR Local POR
['N P01-3 Reading i> 'i S.' Varga R. Wessman L
M. Rushbrook P. Sears V. Nerses-W. Beckner - HLS353 OGC - 15 B18 1
Dennis Hogan:- 3206 MNBB d
E. Jordan - 3701 MNBB G.-Hill (4)-P137 Wanda Jones - 7103 MNBB
-J. Calvo - 11 F23.
ACRS (10) - P-315 GPA/PA - 2 G5 OC/LFMB - MNBi, J. Johnson, Region I 4
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