ML20151D736

Forwards Request for Addl Info Re Chapter 5,advanced LWR Requirements Document.Response to Encl Questions & Answers Requested by 880429

ML20151D736
Person / Time
Issue date:	04/04/1988
From:	Leech P Office of Nuclear Reactor Regulation
To:	Kintner E GENERAL PUBLIC UTILITIES CORP.
References
PROJECT-669A NUDOCS 8804140422
Download: ML20151D736 (10)

Text

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April 4, 1983 Project No. 669 Mr. E.E. Kintner, Chairman ALWR Utility Steering Committee GPU Nuclear Corporation One Upper Pond Road Parsippany, New Jersey 07054

SUBJECT:

SECOND REQUEST FOR ADDITIONAL INFORMATION RELATIVE TO CHAPTER 5, ALWR UTILITY REQUIREMENTS DOCUMENT During the staff's consideration of Chapter 5 of the ALWR Requirements Document, we have determined that additional information is needed in order to complete our review. Enclosure 1 provides questions and comments to which we request your response.

Our current schedule for review of Chapter 5 is based upon receipt of your response by April 29.

If you anticipate that it will be delayed, please inform me so that the schedule can be adjusted accordingly.

Sincerely.

Originni SiF,ned Py Paul H. Leech, Project Manager Standardization and Non-Power Reactor Project Directorate Division of Reactor Projects III, IV, Y and Special Projects Office of Nuclear Reactor Regulation

Enclosure:

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%.....f April 4, 1988 i

Project No. 669 Mr. E.E. Kintner, Chairman i

3 ALWR Utility Steering Comittee l

GPU Nuclear Corporation i

One Upper Pond Road Parsippany, New Jersey 07054 j

SUBJECT:

SECOND REQUEST FOR ADDITIONAL INFORMATION RELATIVE TO CHAPTER 5, ALWR UTILITY REQUIREMENTS DOCUMENT During the staff's consideration of Chapter 5 of the ALWR Requirements Document, we have determined that additional information is needed in order to complete 00* review. Enclosure 1 provides questions and coments to which we request l'

1 your response.

Our current schedule for review of Chapter 5 is based upon receipt of your j

response by April 29.

If you anticipate that it will be delayed, please inform me so that the schedule can be adjusted accordingly.

Sincerely, Paul H. Leech, Project Manager i

3 Standardization and Non-Power Reactor Project Directorate Division of Reactor Projects III, IV, Y and Special Projects Office of Nuclear Reactor Regulation

Enclosure:

As stated cc: J. DeVine, EPRI J. Yedidia EPRI

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REQUEST FOR ADDITIONAL INFORMATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATIVE TO ALWR UTILITY REQUIREMENT 5 DOCUMENT, CHAPTER 5 PROJECT NO 669 MECHANICAL ENGINEERING BRANCH 210.1 The Codes and Standards Rule Section 50.55a of 10 CFR Part 50, requires in Part (g) that pumps and valves which are safety-related be designed to enable the performance of inservice testing of the pumps and valves

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for assessing operational readiness in accordance with Section XI of the Code.

It is an NRC staff position that design of fluid systems that perfern a safety function shall permit pumps and valves to be tested in accordance with Section XI. This nay require installation of adequately sized return lines to test pumps at a point well out on the pump curve, or installation of leak rate testing lines for valves.

Yalves, such as check valves, shall be installed in a manner that permits using system flow conditions to exercise the disc to the open or closed position, or both, as required to perform the safety function. All of the plants which have been licensed by NRC so far have been allowed to request relief from the ASME Section XI in-service testing rules for a limited number of pumps and valves.

These pungs and valves are generally installed in systems in which it is impractical to meet the Section XI rules because of limitations in the system design which make the pump or vaive difficult to test without additional design changes.

Therefore, the staff granted many of these requests for relief because of impracticality or because imposition of these rules would have resulted in hardships to the licensee without a compensating increase in the level of safety.

The underlying reason for the regulation allowing these reliefs from the code was that the detailed piping system designs for all of these plants were completto prior to the time the staff began to implement the ASME Section XI rules.

A plant such as the ALWR, for which the final design is not complete, has 1

sufficient lead time available to include previsions for this type of testing in the detailed design of applicable piping systems.

Therefore, the Utility Requirements Document (URD) should specify that ALWR piping systems will be designed to accommodate the applicable code requirements for inservice testing of purps and valves. With regard to subsequent or future code revisions to the applicable ASME Code for ALWR, requests for relief from certain updated code requirements may still be submitted for staff review in accordance with 10 CFR 50.55a(g).

210.2 The Code permits certain valves to be exercised during cold shutdowns where it is not practical to exercise them during plant operation. These valves are specifically identified by the licensee and they are full-stroke exercised during cold shutdowns. While this i

deviation would be available for the ALWR plant, every attempt should be made in the final design to minimize the number of valves which would require this deviation. The staff requires that the licensee

e 2-provide a technical justification for each valve that cannot be i

exercised quarterly during power operations that clearly explains the difficulties or hazards that would be encountered during that testing. The staff will then verify that it is not practical to exercise those valves and that the testing should be performed during cold shutdowns. Cold shutdown testing of valves identified by the licensee is acceptable when the following conditions are met:

a.

The licensee is to commence testing as soon as the cold shutdown condition is achieved, but no later than 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> after shutdown and continue until complete or the plant is ready to return to power.

b.

Completion of all valve testing is not a prerequisite to return to power.

c.

Any testing not completed during one cold shutdown should be performed during any subsequent cold shutdowns starting from the last test performed at the previous cold shutdown.

d.

For planned cold shutdowns, where ample time is available and testing all the valves identified for the cold shutdown test frequency in the IST program will be accomplished, exceptions to the 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> may be taken.

The URD should specify that the final design of the ALWR plant will confor'm to this staff position for valves tested on a cold shutdown i

frequency, 210.3 When flow through a check valve is used to indicate a full-stroke exercise of the valve disk, the staff position is that verification of the maximum flow rate (identified in any of the plant's safety analyses) through the valve wculd be an adequate demonstration of the full-stroke requirement. Any flow rate less than this will be considered partial-stroke exercising unless it can be shown (by some means such as measurement of the differential pressure across the valve) that the check valve's disk position at the lower flow rate would permit maximum required flow through the valve. This testing must be performed in such a way as to verify the flow through each valve. Thu ALWR plant IST programs should conform to this staff position.

210.4 As part of the resolution of generic issue "Emergency Core Syster.

Design" discussed in Paragraph B.1 of Chapter 5. an increase in the reliability of emergency diesel generator subsystem is being required to reduce the challenge to the ECCS system. The staff position is that the energency diesel generators perform a safety-1 related function.

In order to assure the operational readiness of 4

. the emergency diesel generators subsystem, the appropriate valves in the emergency diesel air starter, cooling water and fuel oil traasfer systems should be included in the IST program and be tested in accordance with the Code. Engine driven pumps are considered to be part of the diesel and need not be tested separately. The URD should require conformance to this position for the ALWR emergency diesel generator subsystem.

210.5 Solenoid operated valves are not exempted from the stroke time measurement requirements of Section XI; their stroke times must be measured and corrective action taken if these times exceed the limiting value of full-stroke time.

The staff will grant relief from the trending requirements of Section XI (Paragraph IWY-3417(a)) for rapid-acting valves; however, in order to obtain this relief, the licensee must assign a maximum limiting stroke time of two seconds to these valves.

210.6 Excess flow check valves perform a safety-related function and should be included in the IST program.

210.7 Control rod drive system valves should also be included in the IST program.

The URD should specify that the final design of the ALWR plant and the IST program will conform to the staff positions noted in items 210.6 and 210.7 above.

210.8 Paragra)h B.5.4 concerning the resolution of generic issues related to "Hig1/ Low Pressure Interface Design" is lacking information on acceptance criteria for leak testing of pressure isolation valves (PIVs). There are essentially two types of P!Vs.

The first type are PIVs whose sizes are small or that are not connected to any accident mitigation systems. The second type are P!Vs whose sizes are of some significance and are connected to an accident mitigation system. Failure of the first type of PIVs would result in a LOCA that should be within the design basis accident and should have no significant impact on the normal plant's ability to shutdown the reactor or to mitigate the consequence of the accident. However, the failure of the second type of PIVs could not only result in a LOCA that is beyond the plant's ability to achieve a normal shutdown of the reactor but also disable part of mitigation system needed to mitigate the LOCA. Therefore, gross failure of the second type of i

PIVs could result in an accident that might be beyond the design accident and might ultimately lead to a core melt.

One purpose of ISTistoensuretheleaktightintegrityofcertainPIVsw1ose failure may impair the plant s ability to shutdown the reactor or to mitigate an accident. The general requirements for leak testing of PIVs are established by 10 CFR 50, Appendix A, GDC 14, 30 and 32, 1

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. 1 which established that PIVs are ASME Section XI Category A valves (seat leakage important). For Category A valves, 10 CFR 50.55a(g) requires that seat leakage testing be conducted on such valves to meet the requirements of ASME Code,Section XI.

For a plant such as an ALWR which will have sufficient lead time to develop its detailed piping design and IST program, the staff re-quires leak rate testing of all P!Ys of the second type noted above.

Please describe the scope of PIY testing, surveillance requirements, and leak rate criteria.

4 210.9 Paragraph ~6.2.2.3 requires that containment isolation provisions be designed to minimize the number of containment isolation valves which are subject to Type C tests of 10 CFR 50 Appendix J.

The staff concurs with this position. Relief from applicable leak test procedures and requirements of Section XI paragraphs IWV-3421 through 3425 for containment isolation valves presents no safety problem since the intent of IWV-3421 through 3425 is met by Appendix J requirements; however, compliance with Paragraphs IWV-3426 and 3427 is considered mandatory.

Those valves, if any, that serve both a pressure boundary isolation function and a containment isolation function must be leak tested to both the Aspendix J and the Section XI requirements. The URD should specify t1at an Applicant is i

expected to meet these requirements for the ALWR containment l

isolation valves.

ELECTRICAL SYSTEMS BRANCH 2

430.1 Paragraph 2.3.3 of Chapter 5 states that, "Systems shall be provided to maintain tie plant in a safe condition during a station blackout, viz., loss of offsite and on-site ac power for eight hours, assuming mechanistic system performance and j

j best estimate analytical methods without a single failure in addition to station blackout." Provide a few examples of how "mechanistic system performance and best estimate analytical methods" might be used in the eight hour station blackout coping analysis. These techniques, that are applied to "risk eval-i uation basis and performance evaluation basis events," may not be appropriate for the station blackout analyses.

Paragraph 1.2.2 of Chapter 5 identifies station blackout as a "risk l

evaluation basis and performance evaluation basis event." "Risk evaluation basis" and "performance evaluation basis" are described as design bases that extend the ALWR design beyond the licensing design i

basis in order to meet the ALWR objectives of increased public safety, improved economics and utility investment protection.

Station blackout, however, may be part of the licensing design basis for nuclear power plants. The Comission will be voting shortly on whether to implement the proposed station blackout rule.

If it is

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I i@lemented, it will be part of the licensing design basis and, l

depending on how and where they are used. "mechanistic system per-f formance and best estimate analytical methodr w y not provide the i

assurance n6eded that a nuclear unit can meet tne requirements of the f

rule.

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l 430.2 With regard to the station blackout 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> coping capability called i

for in Section 2.3.3 of Chapter 5, we note that there is a possi-i bility that this coping duration may not be suffi:ient for all ALWR designs at all plant sites. The draft regulatory guide (Regulatory Guide 1.115) associated with the proposed station blackout rule lists a 16 hour1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> coping duration for a unit that has two redundant emergency power sources with a maximum failure rate per l

demand of.05 and an offsite power source with an expected total loss i

of offsite power frequency (due to grid failures) of equal to or greater than once in 20 site-years.

The actual coping duration would have to be established during the licensing review process for a specific ALWR plant.

430.3 Paragraph 5.3.3.5.2 of Chapter 5 specifies that:

"All controls and instrumentation required for operation of the turbine driven pumps during station blackout shall be i

capable of performing their safety function independent of i

nomal offsite and emergency onsite ac power. Following depletion of the batteries, continued turbine driven pump operation shall be possible through either appropriate failure mode or local manual control."

j Please clarify whether it is intended that the batteries be sized to power the turbine drivei pump loads for the full 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> specified station black-out coping duration, or whether the provision for "failure mode or local manual control" will be used to help achieve the 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> duration. The batteries should be capable of powering these loads for the full 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> coping duration.

j 430.4 Pargraph B.9.3 refers to a "four-or-eight-hour coping requirement" in

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the discussion of the station blackout proposed regulatory requirements.

The staff notes that, in the currently proposed regulatory requirements i

for station blackout, coping durations of 2, 4, 8, or 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> are i

specified.

l 430.5 Paragraph 3.4.5 requires that the engineered safety systems be j

designed such that the onsite power source start time need not be 1

shorter than 20 seconds and the combined start time and load i

sequencing time need not be shorter than approximately 40 seconds.

l The stated ratiotiale for this requirement is that operating plants have sometimes exceeded the current 10 second start time requirement 4

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by a few seconds or have experienced governor stability problems and emergency overspeed shutdowns.

It is not clear, however, how the design of the diesel starting system might be changed to take advantage of this increased starting time and improve the starting reliability of the machine.

If a 1< miter is added to the fuel rack for exam)1e, this might add another failure mode which could actually reduce tie starting reliability. If the longer starting requirement

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is simply used to provide additional margin between the starting of t

the diesel and the first load step, then this could mask a problem with delayed firing of the diesel unless appropriate surveillances l

1 were performed to monitor the delayed firing. There is also a concern i

about how the longer start time would affect the load sequencing intervals and load blocks.

If the longer start time results in shorter load sequencing intervals or larger load blocks, the net reliability of the starting sequence might be reduced.

Please address these issues and provide additional information on how the t

design of the diesel starting system might be changed to take advantage of the increased starting time requirement.

I REACTOR SYSTEMS BRANCH 440.1 Paragraph 5.2.3.5.2 states that: "The valves shall be desigred to ensure that they can be opened if pressure buildup occurs between the valves or inside the valve bonnet as a result of thermal expansion of trapped water."

Does this mean that the valve operators will be sized to open against full reactor pressure?

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I CHEMICAL ENGINEERING BRANCH 450.1 Paragraph 1.2.3.4 What is tha basis for assuming that during a core damage accident release of the fission products from the reactor coolant j

system to the containment will occur in one hour from the time of reactor f

scram?

i 450.2 Paragraph 2.1.3 Core Damage Prevention is shown on Fig.5.2.1, not i

Fig.5.2.2.

450.3 Paragraph 2.4.1.7 The 13 percent limit for hydrogen concentration in the containment does not seem to be conservative:

d)

Sandia has demonstrated experimentally that, for hydrogen-air mixtures at 100 degrees C, detonation can occur at hydrogen j

concentrations as low as 13 percent. No margin of safety is, therefore, provided.

b)

If the hydrogen-air mixture containing 13 percent of hydrogen 4

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deflagrates, the resulting temperature may cause damage to the safety related equipment.

450.4 Paragraph 6.6.2.3 In assessing the containment loads occurring during severe accidents, the thermal and pressure loads generated by these accidents are not added to the other loads such as those caused by seismic events.

What is the basis for this omission?

450.5 Paragraph 8.2.3.9.2 The fouling factor for the heat exchanger in the containment spray system should be 0.001 (Standard T-2.41. Tubular Exchanger ~ Manufacturers Association), correspundng to the value for treated boiler feedwater, rather than 0.0005.

450.6 Paragraph B.8.3 Although the Mark I and Mark II containments are nitrogen inerted and not subject to hydrogen deflagration, under certain circumstances they could become deinerted due to radiolytically generated oxygen.

Was this situation considered in the ALWR analysis?

RISK APPLICATIONS BRANCH 720.1 (a) What is the probabilistic be is for designing ALWRs with only (b) What alternatives two 100% emergancy(diesel generatars?

were considered?

c) Why were alternatives such as four DGs, or two DGs and one or two diverse gas turbine generators rejected?

720.2 (a)

Identify what equipment will be needed to prevent or mitigate severe accidents which will also potentially be subject to environ-ments (e.g. temperature, humidity, pressure, and radiation) beyond the normal design bases for LWRs.

(b) How should the survivability of this equipment be demonstrated?

720.3 Explain how the designs of ALWRs assure that long term containment cooling can be supplied and maintained for months under conditions where a core melt and possibly core-on-the-floor have occurred.

720.4 Is RCIC in the ABWR to be a safety grade system? If not, why not?

720.5 Justify why drpell spray capability should not be AC-independent (a concern especially following station blackout).

720.6 (a) Discuss your mission time for long term cooling following an accident.

(b) Justify the adequacy of the lencth of this time given the potential need for decay heat removal over a several aienth period.

(c) Explain how the ALWR designs assure this cooling can be naintained during this period.

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5 720 7 Are the EFW steam turbines to be self-governing if DC power is lost?

If not, why wasn't this design adopted?

720.8 Coment - Separation of safety from non-safety systems in itself will not resolve the issue of allowable outage times discussed i

in Topic Paper B.6.

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