ML20148D953
| ML20148D953 | |
| Person / Time | |
|---|---|
| Issue date: | 03/18/1988 |
| From: | Leech P Office of Nuclear Reactor Regulation |
| To: | Kintner E ALWR UTILITY STEERING COMMITTEE, GENERAL PUBLIC UTILITIES CORP. |
| References | |
| PROJECT-669A NUDOCS 8803240394 | |
| Download: ML20148D953 (6) | |
Text
__ - _ _
4 March 18, 1988 s
Project No. 669 Mr. E.E. Kintner, Chairman ALWR Utility Steering Comnittee GPU Nuclear Corporation One Upper Pond Road Parsippany, New Jersey 07054
Dear Mr. Kintner:
SUBJECT:
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 some of the questions and comments to which we request your response. A preliminary copy of this enclosure was discussed with Mr. J. Yedidia of EPRI and members of MPR Associates on March 3, 1988.
Additional items will be sent to you shortly.
Our current schedule fcr 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, Original 31g.39 D3 '
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:
As stated cc: J. DeVine, EPRI J. Yedidia, EPRI jdg Distribution:
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March 18, 1988 Project No. 669 Mr. E.E. Kintner, Chairman ALWR Utility Steering Comittee GPU Nuclear Corporation One Upper Pond Road Parsippany, New Jersey 07054
Dear Mr. Kintner:
SUBJECT:
REQUEST FOR ADDITIONAL INFORMATION RELATIVE TO CHAPTr.R 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 some of the questions and comments to which we request your response. A preliminary copy of this enclosure was discussed with Mr. J. Yedidia of EPRI and members cf MPR Associates on March 3,1988.
Additional items will be sent to you shortly.
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, I
Paul H. Leech, Project Manager i
Standardization and Non-Power l
Reactor Project Directorate l
Division of Reactor Projects III, IV, I
V and Special Projects
(
Office of Nuclear Reactor Regulation
Enclosure:
As stated cc: J. DeVine, EPRI l
J. Yedidia, EPRI l
1 l
l
RELUEST FOR ADDITIONAL INFORMATION BY TR OFFICE OF NUCLEAR REACTOR REGULATION RELATIVTfD ALWR UTILITY REQUIREMENTS DOCUMENT, CHAPTER 5 PROJECT NO. 669 PLANT SYSTEMS BRANCH 410.1 Paragraph 2.2.6 indicates that in the design of the decay neat removal systems, heat generation rates from radioactive decay of fission products shall be assumed as equal to ANS Standard 5.1 (October 1979). However, 10 CFP 50.34(g) requires that the aesign be evaluated against the SRP Branch Technical Position ASB 9.2, "Residual Decay Energy for Light Water Reactors for Long Term Cooling." Therefore, provide an evaluation and justification for your proposed approach including a comparison of the AHS standard decay heat generation rates against those of BTP ASP 9.2 to demonstrate how the alterrative proposed provides an acceptable method for complying with the staff requirements.
4:0.2 Paragraph 5.3.2.4 indicates that emergency feedwatcr actuation shall t+ in accordance with the requirement in Paragraph 4.2.3.4 of Chapter 3 which allows the options of automatic or manual initiation of emergency feedwater flow. However, 10 CFR Part 50.34 (f)(2)(xii) requires the provision of autenatic and manual auxiliary feedwater system initiation. Therefore, provide detailed evaluation and/or analysis to justify the deviation.
i 4S0.1 The criteria h,r maximum allowable letkage area for stean bypass of the suppression pool should be addressed.
480.2 raragraph 2.4. ' Mitigation," should also address the requirements of GDC 38 for rapidly reducing the containment pressure following an accident.
In this regard, the sti f position is that contsinment pressure should be reduced to les., thar. 50% of the peak celculated pressure for the design basis loss-of-coolant accident within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> t.fter the postulated accident.
480.3 Criteria for minimum containment pressure analysis for ECCS performance capability have not been addres:e:'.
480.4 Criteria for containment subcompartment analysis and design have l
not been addressed.
l 480.5 With regard to the control of combustible gas concentration inside containment, Sections 2.4 and 6.5, and Paragraph B 8.4.2 inNte that the hydrogen concentration limit should not exceed 13 p m ent and that the mount of hyfrogen released due to metal-water reaction be equivalent to that generated by oxidation of 75 percent of the fuel cladding surrounding the active fuel. These criteria do not meet the guidelines established in Regulatory Guide 1.7 for design basis accidents and the requirenonts specified for severe accidents in 10 CFR Fart 50.34 for the hydrogen control system.
.g.
R.G. 1.7 specifies that the hydrogen concentration limit inside containment should not exceed 4 percent, and that the hidrogen production from metal-water reaction is 5 times the extent of the maximum calculated reaction under 10 CFR Part 50.46 in 2 minutes or that amount that would be evolved from a core-wide average depth of reaction into the original cladding of 0.00023 inches in 2 minutes whichever is greater.
10 CFR Part 50.34(f)(2)(1x)(a) require; that e uniformly distributed hydrogen concentration be maintained in containment which does not exceed 10 percent during and following an accident that releases en equivalent amount of hydrogen to what would be generated from a 100 percent fuel cladding metal-water reaction.
Based on the above criteria, provide a detailed evalaution and/or analysis to justify the identified deviations in your guidelines for hydrogen control.
480.6 Paragraph 6.2.2.1.2 states that if a single isolation valve is employed for an engineered safety feature line (i.e., lines connected to the suppression pool in a BWR and lines connected to the in-containment refueling water storage tank in a PWR), it need not be enclosed in a leak tight enclosure if the line inside containment is submerged under water at all times following a LOCA. This criterion established for containment isolation does not meet the requirements of GDC 56.
In satisfying GDC 56 on another defined basis, Section 3.6 and Note 56.1 in Appendices A and B of ANS 56.2 (1976) provide guidelines which have been found acceptable by the staff for this type of engineered safety feature line. Therefore, Paragraph 6.2.2.1.2 should be revised to conform with these guidelines or justification for the deviation provided.
480.7 Discuss the leak detection capability for lines provided with remote manual containment isolation valves to ensure that adequate information is available to the operator to isolate these lines when required.
480.8 Paragraph 6.2.2.2.3 states that seismic design shall be employed to the extent possible for closed systems outside of containment which penetrate the containment boundary and Type C tests are net required for isolation valves in lines which terminate in closed systems outside containment that are seismical* / designed. These criteria do not meet the requirements of GDC 54, 55, 56 and 57. ANS 56.2 Section 3.6.4 provides the guidelines for these containraent penetrations in order to meet the inter.t of GDC 54, 55, 56 and 57.
These guidelines are:
I a)
The closed systen outside containment is treated as an extension of containment; and b)
The closed system shall be leak tested in accordance with Section 5.3 of the Standard.
Therefore, Paragraph 6.2.2.2.3 should be revised to conform with the i
l ANS 56.2 guidelines or justifiution for the deviation should be provided.
l
" 480.9 Paragraph 6.3.2.1 indicates that containment leak rate testing shall be performed in accordance with regulatory requirements and the test methods shall be in accordance with ANSI /ANS-56.8.
It should be noted that ANSI /ANS-56.8 has not been found acceptable by the staff. A number of unresolved issues were identified by the staff during its review of ANSI /ANS-56.8 Therefore, Paragraph 6.3.2.1 should be revised to reflect the staff's concerns, and guidance should be provided for use by individual referencing applicants on addressing these issues.
489.10 Paragraph 6.3.2.3 states that the maximum interval between Type C tests shall be 30 months rather than the 24-month interval currently required by Appendix J.
The staff noted that Appendix C.1 to Chapter 5 discusses this preposal in terms of risk, occupational exposure, and costs. However, supporting data from operating experience or experiments with an appropriate analysis should also be provided to justify this deviation from Appendix J requirements.
INSTRUMENTATION AND CONTROL SYSTEMS BRANCH 420.1 Paragraph 2.3.2 includes a requirement that "Equipment, piping, and electrical raceways shall be located, employing special separation to the extent practical - -
." Please note that the minimum separations stated in Regulatory Guide 1.75 must be met.
420.2 Paragraph 8.2.3.13.2 states that "The CSS shall be automatically actuated." The availability of manual actuation should also be required, per IEEE-279.
RADIATION PROTECTION BRANCH 471.1 In regard to source terms, EPRI is proposing to use a realistic fission prcduct source term, rather than a non-r chanistic source term, as the licensing design basis for the ALWR.
The principal elements that EPRI proposed for the realistic source term are: (1) not requiring containment spray additives; (2) not requiring activated chcrecal filters in relec e pr.chs; (3) allowing credit for radionuclide scrubbing by BWR suppression pcols; (4) not requiring )
an instantaneous release of fission products to the containment; (5 l
using realistic containnent leak rates; and (S) increasing the sllowable containment leak rate to greater than er equal to I
l 0.5%/ day. Presumably, one effect of using a realistic source term is that, all other factors remaining the same, the exclusion area, low pcpulation zone, and population center distance for an ALWR would be less than for an LWR.
a.
Describe the eff ect of using a realistic source term on the size of the exclusion area, low popula: ion zone, and population l
center distance (i.e., see 10 CFR 100.11). Be quantitative to the extent feasible.
1 I
l
4 b.
Is EPRI proposing any minimum distances for the exclusion area, low population zor.2 and population center, or would these distances be determined at the time a specific license application is submitted?
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