ML18031A026

Requests Addl Info Re Seismic Classification & Sys Quality Group Classification in Order to Complete Review of Appl Oper Lic

ML18031A026
Person / Time
Site:	Susquehanna
Issue date:	11/07/1978
From:	Parr O Office of Nuclear Reactor Regulation
To:	Curtis N PENNSYLVANIA POWER & LIGHT CO.
References
NUDOCS 7811150029
Download: ML18031A026 (63)

Text

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Docket File NRC PDR Local PDR LWR P3. File Boyd Docket Nos 50-and

-388 osure:

Vassallo Williams Parr Miner Rushbrook Mattson Ross Knight R. Tedesco R.

DeYoung V. Moore R. Vollmer

- M. Ernst R. Denise R. Hartfield OELD IE (3)

C. Ferrell M. Fliegal B. Belke J. Boegli B. Kirkwood Mr; =Nortttan W. Curtfs Yfce Presfdent - Engfneerfng and Constructfon Pennsylvanfa Power and Lfght Company 2 North Nfnth Street Allentown Pennsylvanfa 18101

Dear Mr. Curtfs:

BCC:

JBuchanan TAbernathy ACRS (16)

SUBJECT:

SUSlgEHANNA STEAM ELECTRIC STATION UNIT NOS. 1'ND 2-RE(UEST FOR ADDITIONAL'.INFORMATION As, a result of our revfew of your.applfcatfon for operating licenses for the-Susquehanna Steam Electric; Plant we ffnd that we need addftfonal information concerning Codes and Standards Rule (10 CFR Part 50 Sectfon 50.55a),

Sefsmfc Classfffcatfon and System gualfty 6roup Glassfffcatfon.

The specfffc information requfred: fs lfsted fn the Enclosure.

Please fnform us.of the date when this requested addftfonal fnformatfon.

wfll'e avaflabkh i'or our revfew.

Please. contact us ffyou desire any dfscussfon or clarification of the fnformatfon requested.

Sfncerely, Prldnal Slgaed 5g,~

Naa Parr Olan D

Panr, Chief Lfght Mater Reactors Branch No.

3 Dfvfsfon of Prospect Management Enclosure's. 3tated,-

Cc w/enclosure:

See next page 78'Ill~a~~~@ >

ORRICC~

SURHAMC+

LWR-83:LP LWR 83:

C SMiner.:ctv' OParr OATCW ll/ Cci /78 11/'

/78 NRC FORM 518 (9-76) NRCM OXEO

• UIS 4OVCRNMCNT RRIHTIH4 ORFICCs ISTS 4~~

Hr. Norman W. Curtis

. cc:

Hr. Earle H. Mead Project Manager Pennsylvania Power 8 Light Company 2 North Ninth Street Allentown, Pennsylvania 18101 Jay Si lberg, Esq.

Shaw, Pittman, Potts 5

Trowbridge 1800 H Street, N.

W.

Washington, D. C.

20036 Hr. William E., Barberich, Nuclear Licensing Group Supervisor Pennsylvania Power

& Light Company 2 North Ninth Street Allentown, Pennsylvania 18101 Edward H. Nagel, Esquire General Counsel and Secretary Pennsylvania Power 5 Light Company 2 North Ninth Street Allentown, Pennsylvania 18101 Bryan Snapp, Esq.

Pennsylvania Power 5 Light Company 901 Hamilton Street Allentown, Pennsylvania 18101 Robert H. Gallo Resident Inspector P. 0.

Box 52 Shickshinny, Pennsylvania 18655

ENCLOSURE RE UEST FOR ADDITIONAL INFORMATION SUS UEHANNA STEAM ELECTRIC STATION DOCKET NOS.

50-387 ANO 50-388

005 0

AUXILIARYSYSTEMS BRANCH 005.1 The statement in Section 5.2.1.1 of the FSAR with regard to your com-pliance with 10 CFR Part 50, Section 50.55a, Codes and Standards

Rule, is incorrect as a number of guality Group A components within the reactor coolant pressure boundary are not in conformance with the applicable ASME Boiler and Pressure Vessel code and addenda as required by the rule.

In Amendment 13 to the Susquehanna Steam Electric Stations'SAR and in your letter ER 100450,. File 040-2, received by the Staff on March 1, 1974, you provided an analysis of anticipated deviations from the codes and standards rule requirements set forth in the provisions of Section 50.55a, 10 CFR Part 50, based on a Construc-tion Permit Date of November 2, 1973, for the Susquehanna reactor pressure

vessels, reactor recirculation piping, reactor recircula-ti,on system

pumps, main steam line isolation valves, and main steam safety/relief valve...

Based on this information and on certain additional commitments relative to the reactor pressure

vessels, the AEC in a letter dated June 20, 1974, in accordance with paragraph 50.55a (a)(2)(ii), granted approval for relief from the rule for these components and acceptance of the ASME Section III Code and Addenda specified in Amendment 13 to the PSAR and letter ER 100450, File 040-2.

005-2 Revise Section 5.F 1.1 of the FSAR to correctly reflect the status of each guality Group A component within the reactor coolant pressure boundary.

005.2 In Table 3.2-1 of the FSAR identify the applicable principal construction codes and standards in those cases where this information is now missing throughout the table.

005.3 The 831.1 component code identified in Table 3.2 of the FSAR for the diesel lube oil system piping and valves is inconsistent with the guality Group C (Safety Class

3) classification for these components.

The diesel generator lubrication system piping is also identified in Section 9.5.7.1 of the FSAR as designed in accordance with ASME Section III, Class 3.

Resolve this inconsistency and revise the FSAR as appropriate.

005.4 The quality group (safety class) classification, seismic classification, component code and quality assurance requirements for the components of the Emergency Service Water System have been omitted from Table 3.2-1.

Revise Table 3.2-1 to include this information.

005.5 The quality group (safety class) classification, seismic classification, component code and quality assurance requirements for the spray pond system piping has been omitted from Table 3.2-1.

Revise Table 3.2-1 to include this information.

0 005-3 Verify that all components within the reactor coolant pressure boundary as defined in 10 CFR Part 50.2(v) are classified guality Group A in compliance with the Codes and Standards Rule, Section 50.55a of 10 CFR Part 50, or as a minimum, are classified guality Group B if the components meet the exclusion requirements of the rule.

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Distributi Docket Fil NRC PDR Local PDR LWR 83 File R.

Boyd Docket Hos.

50-38 w/e D.

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Vassallo Williams Parr Miner Rushbrook Mattson Ros's J.

Knight R. Tedesco R.

DeYoung V. Moore R. Vollmer M. Ernst R. Denise R. Hartfield C. Ferrell OELD M. Fliegal IE (3)

B. Belke J. Boegli BCC:

JBuchanan TAbernathy ACRS (16)

,.DCT 31 QQ Mr. Norman >t. Curtfs Vfce President - Engineering and Construction Pennsylvania Power and Light Company 2 North Ninth Street Allentown, Pennsylvania 18101

SUBJECT:

SUSQUEHAfiHA STEAf't ELECTRIC STATION UNIT NOS.

1 AND 2-

, 'EQUEST FOR ADDITIONAL INFORIlATION As a result of our review of your application for'perating licenses for the Susquehanna Steam Electric'Plant we find that we need additional information in the areas of Accident Analysis, Effluent Treatment, Hydrology and Quality Assurance.

The specific information required is listed fn the Enclosure.

Please inform us of the date when this requested additional information will be available for our review.

Please contact us ff you desire any discussion or clarification of the information requested.

Sincerely, Originak Slggef by, 0 Q.tea.

Dian D. Parr, Chief Light tfater Reactors Branch Ho.

3 Division of Project Management

Enclosure:

As Stated cc w/enclosure:

See next page oaa!CE~

SURNAME~

Ll<R $3~P St1iner/LL 10/p~ /78

, LI!R II3 BB

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/78 NRC FORM 318 (9-76) NRCM 0240 N UI 0, OOVERNMENT PRINTINO OPPICEI 'ISTS 02(402@

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Hr. Norman W. Curtis OCT 31i978 cc:

Hr. Earle M. Mead Project Manager Pennsylvania Power 8 Light Company 2 North Ninth Street Allentown, Pennsylvania 18101 Jay Silberg, Esq.

Shaw, Pittman, Potts 5

Trowbridge 1800 H Street, N.

W.

Washington, D. C.

20036 Hr. William E. Barberich, Nuclear Licensing Group Supervisor Pennsylvania Power 8 Light Company 2 North Ninth Street Allentown, Pennsylvania 18101 Edward H. Nagel, Esquire General Counsel and Secretary Pennsylvania Power 8 Light Company 2 North Ninth Street Allentown, Pennsylvania 18101 Bryan Snapp, fsq.

Pennsylvania Power 8 Light Company 901 Hamilton Street Allentown, Pennsylvania 18101 Robert H. Gallo Resident Inspector P. 0.

Box 52 Shickshinny, Pennsylvania 18655

ENCLOSURE RE UEST FOR ADDITIONAL INFORMATION SUS UEHANNA STEAM ELECTRIC STATION DOCKET NOS.

50-387 AND 50-388

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312. 0 312.11 (2.i.l) 312.12 (2.1.1)

ACCI'DENT 'ANALYSIS'BR NCH.

Table 2.2-1 insthe PS+, qtggejgtlgt<ctwo3p jje].jiiesi<j'rf-'the1C'iniety qo f. sjte-'@'"e's edr foj

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None of the maps in the FSAR clearly show the exclusion area boundary.

Provide a full scale section of the USGS map of the Berwick, Pa.

quadrangle which clearly shows the exclusion area as well as the plant boundarY.

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'~provide suffjcien'0-. detai~l"':}>>':

312. 13 (2.1.1)

Although it is not mentioned in. FSAR Section 2.1.3.4 Low Popu-lation Zone, Figure 2.1-1 shows a,race track or an athletic field approximately 1 1/4 miles southwest of the reactor site.

Indicate the use of this facility, the peak attendance and fre-quency of use.

312.14 (6.i.2)

It is stated that you will comply with ANSI N101.2.

What is your intended degree of compliance with Regulatory Guide 1.54, "equality Assurance Requirements for Protective Coatings Applied to Mater-Cooled Nuclear Power Plants?"

If there are any coat-ing materials not qualified according to Reg.

Guide 1.54, provide estimates of their quantities to show that these quantities are insignificant.

312.15 (15.6.5) 312.16 (15'.6.5)

In reference to question 021.30, provide a graph which shows the secondary containment pressure following a loss of. coolant accident during the switch over from the normal ventilation system. exhaust to operation of the standby gas treatment system.

Please. indicate the length of main steam line between the outboard and'. inboard MSIVs.

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321.0 Effluent Treatment S stems Branch 321.6 Your response.to guestion 321.5 on the solidification process control program and the parameters to be considered for the solidification of waste is not adequate.

In accor'dance with BTP-ETSB 11-3, provide more detai 1 concerning the process control program including the following:

< (1),'ata concerning the expected waste types to be processed.

The process control program should be based on tests performed with simulated waste formulations based on the expected inputs.

You should discuss how the, process control program considers the chemical constituents of the'aste

stream, the pH of the waste

stream, boric acid content, solids content of the waste, concen-tration and type of radwaste, curing time, etc.

~ (2)'ata concerning the solidification agents (cement

+ silicate') to waste ratios to be used.

The process control program should con-sider the correct ratios for the various input types and contamin-ant levels.

g (3)>Data concerning the effects of various contaminants on the solidi-fication process.

Specifically, address oil and detergent content in wastes, lab chemicals, and non-depleted ion-exchange resins.

,>(4))Discuss the experimental procedures to be used in your process I II Oi 1i f

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Solid Radwaste System as it relates to your process control program to assure a satisfactory solidified product.

Where will the waste be sampled?

Discuss how the results of the process control program will be analyzed and used as operational considerations.

(5)

We are not familiar with the material, "Safety Set.".

Provide a

product description, including the chemical or physical method of solidifying surface liquid during expected process conditions.

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371. 0 371.19 (2.4.1)

HYDROLOGY -

METEOROLOGY BRANCH -

HYDROLOGY SECTION Provide a map of the site clearly showing the topography a~ altered by ttg plant.

Note that CESAR:Figure= 2.'4-1; is. inadequate-.:because

.it is

'very difficult to see'h'e contours-. in'he vicin'sty,';of,. the plant.'-'-

371. 20 (2.4.2)

Describe the "pressure resisting doors" used to prevent water from reachina safety-related equipment.

Document that they are 4

water tight for the maximum water level they must withstand.

Indicate what.procedures will be used to ensure that the doors will be properly closed durina a flood.

Alternately, if you can document that the maximum water level will be below the sill level of the doors to all safety-related buildings, it may not be necessary to keep the doors shut.

371.21, (2.4.8)

You state, on page 2.4-29,',, of the'SAR~,~ that "".-.v.all 'safety-.rel'ated equipment,[in the" ESSW pu'mphouse]

are located iat high'er"'elevation ~[than the,'84.7 feet MSL you calculated as the maximum wind wave runup 3

and has suitable protection."

What is the elevation of the safety-related equipment and what is the suitable protection?

371.22 Please provide a copy of, or a better reference to the TANS report (2.4.>>)

referred to in your response to 9371.6.

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'71.23 You state, on page 2.4-39'of,.the FSAR,'hat'he river>low level<alarm is set at 488.5 feet NSL.

From the stage discharge curve)FSAR..Figure."~2.4.6, 7"

that level corresponds to a discharge of about 5000 cfs.

From the l

discharge-duration curve FSAR'Figure 2'.4-'30 the- 'river"discha'rge is below 5000 cfs abo~t 40 percent of the time at Wilkes-Barre.

Since the discharge-duration relationship at the site would not be very different than at Wilkes-Barre, it appears that the low level alarm would be activated quite often.

What is the purpose of the alarm and what happens when it is activated?

371.24 Indicate how you intend to ensure spray pond cooling capability (9.2.7) beyond 30 days, especially if:

j (1) ~the Susquehanna River flow is below the level at which you can withdraw water in compliance with 18CFR Part 803.

'j(2) )the river stage is below ii'Ip.c ne..ded for the intake I

system to operate.

We note that on page 9.2-26, ofl;the FSAR~oyou~'refer)to'Section,;1'3.3 which in'urn refers to your emergency plan.

We were unable to find a discussion of makeup water to the spray ponds in that document.

C 371.25 'ou state, on.FSAR-page 9;2'-26, that. at'>> tim'est'of~subfreezing temoeratures return flow to the spray pond will be first discharged directly W

into the pond, through a by-pass line, without passing through the spray network.

Please

indicate, on a diagram of the pond,

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the location of the by-pass line and document that its location precludes short ciriuiting of hot water to the intake without significantly thawing the pond.

Document that the return temperature will rdmain below the design maximum temperature at all times.

371.26 (9.2.7)

On page 9.2-34 "o'fiuthe,'FSAR ~you refer,.'to, a', Appen'dix;D'; which".-have not b'een able to find in the FSAR.

Please either direct us to its location in the FSAR, or if not in the FSAR, provide the document.

371.27 Model studies, performed during the Construction Permit i(CP) review, indicated that the spray ponds, as designed, would be capable. of providing cooling water at a temperature below the design maximum for the shutdown of both units during conditions specified in Regulatory Guide 1.27.

The ability of the as built spray ponds to meet the design bases adopted at the CP must be confirmed by actual performance tests.

I Specifically, tests to confirm that the pond responds in a manner consistent with the model studies previously used to estimate pond performance, are needed.

Commit'o provide a detailed description of your test plan, procedures and analyses techniques for NRC staff review and approval prior to operation of Unit 1.

The plan;-should 'ecognize the availability of heat from Unit 1.

Your schedule for the tests and analyses should allow for NRC staff review and approval prior to 1oadinn fuel for Unit 2.

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QUALITY'SSURANCE - OPERATIONS.

The Quality Assurance Branch (QAB) has reviewed Pennsylvania Power and Light Company's (PP&L) Fire Protection Report (dated January 18, 1978) for Susquehanna Steam Electric Station (SSES)

Units 1

and 2.

This report was submitted in response to Mr. Boyd's letter of September 30, 1976.

Based on our review of this information, we find that adequate information has not been submitted by PP&L to permit completion of the QAB review of the fire protection program.

Item 26 (pg. 3-48) of your submittal does not indicate what the management control of the QA organization consists of.

The description for QA management should consist of ('1) formulating and/or verifying that the fire protection QA program incorpo-rates suitable requirements and is acceptable to the management responsible for fire protection through review, surveillance, and audits.

Performance of other QA program functions for meeting the fire protection program requirements may be per-formed by personnel outside of the QA organization.

The QA program for fire protection should be part of the overall plant QA program..

These QA criteria apply to those items within the scope of the fire protection

program, such as fire protection systems, emergency lighting, communication and breathing apparatus, as well as the fire protection require-

.ments of applicable safety-related equipment.

We find that your response to Mr. Boyd's letter of September 30,

1976, does not describe sufficient detail to address the ten specific quality assurance criteria in Branch Technical Position ASB 9.5-1.

In order fop the QAB to fully evaluate your approach for meeting these criteria, additional detailed description is necessary.

Examples of the detail we would expect PP&L to consider are provided in Attachment 6 of Mr. D.

B. Vassallo's letter of August 29, 1977.

If, however, you choose not to provide this detail, you may apply the same controls to each criterion that are commensurate with the controls described in your QA program for operations.

These controls would apply to the remaining construction activities and for the operations phase of Unit Nos.

1 and 2. If you select this method, a statement to this effect would be adequate for our review of the fire protection QA program.

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421-2 421. 3 (17.2.1) 421. 4 (17.2.1)

Provide a description of how the QA Supervisor (located onsite) communicates with the offsite QA organizations relative'to matters concerning QA/QC, and describe those conditions for determining'when these actions should take place.

The offsite/

onsite interface should also be shown on the applicable organizational charts in,the QA program description.

Identify on organizational charts the reporting relationship of the Nuclear Review Board.

421.5 FSAR>> Figure-,17.2-2 has'an", organizational,"block listed's<,"others."

(17.2.1)

Clarify what "others" are and describe their QA/QC functions, if any.

421. 6 (i7.2.i) 421. 7 (17.2.1) 421. 8 (i7.2.i )

421. 9 (17.2.i) 421.10 (17.2.i )

421.11 (i7.2.1) 421.12 (17.2.1)

Describe in more detail the specific responsibilities of the Nuclear Quality Assurance Staff in executing the SSES QA program.

Describe in more detail those "quality activities" (ref, FSAR page 17.2-6) performed by the Manager, Power Production.

Describe provisions which assure that the Vice-President, Systems Power and Engineering, maintains a continuing involvement in QA matters and how he cormunicates through intermediate levels of management.

(e.g., review and concurrence of SSES operations, administrative control, and operational QA program).

Clearly identify the individual/position responsible for having overall responsibility and authority for the SSES operational QA program.

Describe the amount of nuclear quality assurance experience required for the position of Quality Assurance Manager.

The amount of experience should be at least equal to the one year experience listed in paragraph 4.4.5 of ANSI/ANS-3.1-1978, "Selection and Training of Nuclear Power Plant Personnel."

Describe the qualifications established for the QA Supervisor "regarding quality assurance and quality control related experience.

Describe measures which assure that personnel (including those outside the QA/QC organization) performing QA/QC functions have sufficient authority and organizational freedom to:

4 a)

Iden tify qual'i ty problems.

b)

Initiate, recommend, or provide solutions through designated

channels, and c)

Verify implementation of solutions.

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-421-3 This description should also include measures to assure that verification of conformance to established requirements is accomplished by individuals or groups who do not have direct responsibility for performing the work being verified.

421.13 (17.2.1) 421.14 (17.2.2)

Clarify whether'he stop work authority vested i n the Hanager

- NgA is delineated in writing.

Descr'i;be. provis.ions which assure that management (i.e.,

above or 'outside the gA organizati'on) annually assesses the scope, status, implementation, and effectiveness of the gA program to assure that the program is functioning adequately and compl ies with 10 CFR Part 50, Appendix B

criteria, and that the results of this assessment are documented.

421.15 (17.2.2) 421.16 (17.2.2)

Table 17.2-1 of the FSAR addresses those Regulatory Guides and ANSI standards applicable to the operational gA program and the degree of compliance thereto.

Since the docketing of your application (July 31, 1978), certain of these Regulatory Guides (RG) ',and ANSI standards have been upgraded and differ from the dates stated in Table 17.2-1.

Therefore, update your application, and provide a specific commitment to comply with the regulatory positions of each of the following'egulatory Guides and ANSI standards:

RG 1.23, Rev.l; RG 1;33, ReV 3'G;;1.38, Rev.

2; RG 1.39, Rev. 2; RG 1.116, Rev. 0-R; RG 1.123, Rev. 1; and ANSI N45.2.12, Draft 3, Re'v. 4, 2/22/74 or ANSI N45.2.12,. Draft 4, Rev.

2, 1/1/76, as supplemented by regulatory position 4 of Regulatory Guide 1.33, Rev.

2 (2/78).

Any exceptions and/or alternatives to the above Regulatory Guides/ANSI standards should be described

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in sufficient supporting detail to allow for NRC evaluation and acceptance.

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It is not clear as to your interpretation of the term "Commitment to the extent required by ANSI N18.7-1976" as used in,'FSARG Table 17.2:.l. I

,iP1eas'e'p'rovide

'a more detailed 'explan'ation~of what- "Commitment to ith'e

'exte'nt-'required~ be 'ANSI" N18. 7-1976" means'o'PAI 'and h'ow'it: i s'< to". be

.'used. to-assured'onsistent inter'pretation within, PP8L.

421.17 (17.2.2) 421.18 (17.2.2)

Describe those provisions which assure that the docketed gA program description, particularly the commitment to Regulatory Guides and ANSI standards, will be properly carried out and with the use of gA procedures.

Provide a summary description on how responsibilities and control of quality-related activities are transferred between PP8L and principal contractors during the phaseout of design and construction and during preoperational testing and plant turnover.

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-421-4 421.19 (17.2.2) 421.20 (17.2.2) 421. 21 (17.2.7) 1 Describe measures to assure that appropriate,'10 CFR"Pa'r'tr50~'Appendix B

f requirements,will<<:be:applied toithe.preoperational'<test program.

Describe provisions which assure that the NRC will be notified of changes to the accepted SAR gA program description prior to implemen-tation and of',changes to organizational elements within 30 days after announcement.

(Note - minor editorial changes or personnel reassignments of a nonsubstantive nature do not require NRC notification.)

r Identify those individuals evaluating the suppliers'apabilities to provide acceptable quality services and products prior to the 'award of procurement order or contract.

(gA and Engineering should participate in the evaluation of those suppliers providing critical components.)

421.22 Clarify whether the purchase of spare or replacement parts of safety-(17.2.7) related structures,

systems, and components are subject to controls at least equivalent to those used for the original equipment.

421.23 Describe measures which assure that records are identifiable and (17.2.17) retrievable.

421.24 Describe provisions to assure that the "offsite" gA organization:

(17.2.18) a.

Conducts sufficient audits to verify the activities conducted by the "onsite" gA organization.

b.

Reviews and concurs in the schedule and scope of audits performed by the onsite gA organization.

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-38 NRC PDR Local PDR LWR k3 File R.

Boyd D. Vassallo F. Williams

0. Par'r Mr. Norman W. Curtis Vice President -'ngineering and Construction Pennsylvania Power and Light Company 2 North Ninth Street Allentown, Pennsylvania 18101 Miher Rushbrook Mattson Ross Knight Tedesco DeYoung Moore Vollmer OCT 18 3978 M. Ernst R. Denise R. Hartfield OEL'D IE (3)

BCC:

JBuchanan TAbernathy ACRS (16)

Dear lh'. -Curtis:

SUBJECT:

SUSQUEHANNA STENf ELECTRIC STATION UNITS HOS.

1 ND 2 REQUEST FOR ADDITIONAL IHFOMfATIOH As a result of our review of your application for operating licenses for the Susquehanna Steam Electric Plant we find that we need additional information in the area of training programs and procedures.

The specific information required is listed in the Enclosure.

Please inform us within 10 days after receipt of this letter of the date when this requested additional information will be available for our review.

Please contact us ifyou desire any discussion or clarification of the information requested.

Sincerely, Original Signa/ by O. D. Para Olan D. Parr, Chief Light'Water Reactors Branch Ho.

3 Division of Project Management

Enclosure:

As Stated cc w/enclosure:

See next page LWR P3:LPM SMiner/LL OPPICR~

SURNAME~

LWR k3:BC

---@1fS$7 ODParr OATR~

10/~~ /78 10/ U /78 NRC FORM 318 (9.76) NRCM 0240 A Ul 0, OOVSRNMRNT PRINTINO OIIFICRI IOTO 020 023

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41 CP v,r

Mr. Norman W. Curtis OCT 13

<978 cc:

Mr. Earl.e.M.

Mead Project Manager Pennsylvania Power

& Light Company 2 North Ninth Street Allentown, Pennsylvania 18101 Jay Silberg, Esq.

Shaw, Pittman, Potts Trowbridge 1800 M Street, N.

W.

Washington, D. C.

20036 Mr. William E. Barberich, Nuclear Licensing Group Supervisor Pennsylvania Power

& Light Company 2 North Ninth Street Allentown, Pennsylvania 18101 Edward M. Nagel,'squire General Counsel and Secretary Pennsylvania Power

& Light Company 2 North Hinth Street Allentown, Pennsylvania 18101 Bryan Snapp, Esq.

Pennsylvania Power

& Light Company 901 Hamilton Street Allentown, Pennsylvania 18101 Robert M. Gallo Resident Inspector P. 0.

Box 52 Shickshinny, Pennsylvania 18655

ENCL'OSURE RE UEST FOR ADDITIONAL INFORMATION SUS UEHANNA STEAM ELECTRIC STATION DOCKET.,'!NOS'3;".50-'87."AND"50-388

44,1'I.O

'OPERATOR'L'ICEI'1SING'BRANCH - TRAINING

.'SARrFijure-13.2-1',".Revision.:1;,.'shows that:-16"people will~.

'i 'i be licensed prior to fuel loading.

Provide the number of people who will be trained in your licensed operator training program.

This number. should not only meet Technical.Specification requirements but should also I

allow for examination contingencies and avoidance of j

planned overtime during the startup phase.

We recom-ment the training of at least 25 people.

What are your plans for additional training

--, in the event that fuel loading is substantially delayed?

An acceptable method of maintaining the required level of training if fuel loading is significantly delayed would be to initiate the requalification program.

tFSAR'ir'ure-l3.2-1",'evision~~l",cindica'tes that a<Pre-.-,license

~

Refresher Course will be conducted six months prior to fuel loading.

We consider it highly desirable that license applicants participate in a short sim-ulator course immediately prior to the examinations.

Is it PP8L's intention to provide such a course?

I It@

441-2 441.4 (13.2. 1. 1. 2. 1)

State the methods used to evaluate the training pro-gram effectiveness for Phase I, Phase II, Phase V,

and Phase VI training.

441e5 (13.2.1.1.2.1)

The BMR simulator course is taught at the General Electric Training Center in Morris, Illinois.

Our position is that individuals seeking licenses for the Susquehanna Plant will have to participate in train-ing programs that utilize a Nuclenet simulator, if such a simulator is operational.

Provide a commit-ment that simulator training will be conducted at a Nuclenet simulator, if operational prior to fuel

loading, and identify the simulator to be used for this training.

441. 6 (13.2.1.1e7)

The Susquehanna Fire Safety Training program is un-acceptable.

Provide a detailed description of the fire protection training and retraining for the crit-ical plant staff and replacement personnel which meets the following acceptance criteria:

A.

Fire Bri ade Trainin (1) Instruction

, (g) Instruction in all the topics listed in, (d) below'( should be administered to individuals prior to assignment as a fire brigade member.

i,,'(b)'efresher instruction should be providedIbo a11 fire brig-age members on a regularly scheduled basis of not less than four sessions a year.

The sessions shall be repeated at a

frequency of not more than 2 years.

441-3

'(c)J The instruction shall be provided by qualified individ-

'als, knowledgeable and experienced in fighting the types of fires that could occur in the plant and in using the types of equipment available in a nuclear power plant.

Members of the Fire Protection Staff and fire brigade leaders may conduct this training.

,'(d),'The scope of the instruction should include the following items:

(i)

An identification of the fire hazards and associated types of fires that could occur in the plant, and an identification of the location of the hazards, includ-ing areas where breathing apparatus is required,

'egardless of the size of the fire.

(ii) Identification of the location of installed and port-able fire fighting equipment in each area, and famil-iarization with layout of the plant including access and egress routes to each area.

(iii)

The proper use of available equipment, and the correct method of fighting each type of fire.

The types of fires covered should include electrical fires, fires in'cables and cable trays, hydrogen fires, flammable liquids, waste/debris fires, and record file fires.-

(iv)

Indoctrination in the plant fire fighting plan, with coverage of each individual s responsibilities, includ-ing changes 'thereto.

(v)

The proper use of breathing equipment, communication, lighting and portable vent'ilation equipment.

(vi)

A detailed review of the procedures, with particular emphasis on what equipment must be used in particular areas.

(vii)

A review of latest modifications, additions or changes to the facility, procedures, fire fighting equipment or-fire fighting plan.

(viii). The proper method of fighting fires inside buildings and tunnels.

,~ ('e) In addition, special instruction should be provided for fire brigade leaders in directing and coordinating fire fighting activities.

441=4 ru (2)

Practi ce Fract)ce sessions should be held for fire brigade members on the proper method of fighting various types of fires.

These sessions should provide brigade members with practice in extinguishing actual fires, except in the case of ener-gized cables.

Practice sessions should be conducted at facilities sufficiently remote from the nuclear power plaht so as not to endanger safety-related equipment.

These practice sessions should be provided at regular intervals, but not to exceed one (1) year.

'ractice sessions should also be conducted that require fire brigade members to don protective equipment, including emer-gency breathing apparatus.

These practice sessions need not include fire fighting.

These practice sessions should be

'rovided at regular intervals, but not to exceed one (1) year.

'.3)

Dril1 s 7ireeerigade drills should be performed in the plant so that a fire brigade can practice as a team.

Drills should include the following.

(a),The simulated use of equipment for the various situations and types of fires which could 'reasonably occur in each safety-related area.

e (b) Conformance, where possible, to the established plant fire fighting plans.

'c)'Operating fire fighting equipment where practical.

This would also include self-contained breathing apparatus, comnunication equipment and portable and/or installed ventilation equipment.

~ (d) The drills should be performed at regular intervals, but not to exceed three months for each fire brigade.

The minimum number of fire brigade drills conducted within a period of three months shall be equal to the number of operatinq shifts at the station.

Each individual member of the fire brigades shall participate in at least two drills per year.

At least one drill per year for each fire brigade shall be unannounced.

(e) Periodically (at least annually),

these drills should include off-site fire department personnel.

These drills should also conform with the facility plan for coordina-tion with off-site fire departments.

~(f),~The drills should be preplanned to establish the train-inq objectives of the drills.

The dri'lls should be critiqued to determine how well the training objectives.

have been met.

At a mini'mum, the critique should assess:

(i)

Fire alam effectiveness, response time, selection, placement and use of equipment.

(ii)

The leader's direction of the effort and each member's response.

B.

Other'Station Em lo'ees P(l)! Instruction

'l (a)'lInstruction shall be provided for all employees once a

year.

It shall be repeated on an annual basi's.

The instruc-tion shall be given, as appropriate, on (i) the fire pro-tection plan.(ii)'.evacuation routes, and(".(-iii),procedure for reporting a fire.

'<<(b)'nstruction shall be'provided for security personnel that addresses (ij) entry procedures for outside fire departments

~ (ii):crowd..control

(( forpeo'pl e-,exiting;,the.; statjoq,- and (iii )'<

procedures for reporting potential fire hazards observed when touring the facility.

(c) Instruction should be'provided to all shift personnel that complements that provided members of the fire brig-ade.

< (d) Instruction shall be provided to temporary employees so that they are familiar with (i) evacuation signa1s,;.(ii) evacuation routes and,(iii) procedure,,for', reoorting fires.

((2) 'Dii'lie All employees-should participate in an annual evacuation drill.

C.

Fire Protection'Staff Training for the fire protection staff members inc'1ude courses in:

A 0} design and maintenance of fire detection, suppression and ex-tinguishing systems,

,,(2);fire prevention techniques and procedures,

}I l

IJ r'l~

~

441-6 (3) training and manual firefiohti'ng techniques and procedures for plant personnel and the fire brigade.

D.

Off-Site Fire'De artments Training for the off-site fire departments include courses in basic radiation principles and practices, typical radiation hazards that may be encountered when fighting fires and related procedures.

E.

Constiuction Personnel Training for construction personnel should include instruction in reporting fires, alarm responses and evacuation routes.

441.7

( 13.2.2. 1. 3)

Revise paragraph 13.2.2.1.3 in,the FSAR:to-indicate the following (1)',An individual who prepares, administers, or grades a written examination need not take the examination.

A maximum of three licensed personnel may be exempted under this condition.

(2) 'Retraining lectures may use training aids such as video tapes and films in lieu of an instructor.

However, no more than 505 of the lectures may be supplemented by use of training aids.

441. 8 (13.2.2.1.4)

Oral exams are acceptable for determining whether or not an individual resumes licensed duties after re-ceiving a grade of 705..or less on an annual exam.

However, the individual must remain on a Performance Review Program until a grade of 705 or better is obtained on a written examination.

-441-7 441. 9 (13.2.2.2)

As a minimum, refresher instruction on administrative, radiation protection, emergency, and security procedures should be provided to all non-licensed personnel.

442.1 (13.5.1.3)

Provide a diagram of the control area that indicates the area designated "at the controls".

442.2 (13.5 1.3)

The description of administrative procedures should in-clude Fire Protection and Temporary Procedures of an administrative nature.

442.0, 0 erator Licensin Branch Procedures 442.1 (13.5.1.3)

Provide a diagram of the, control area that indicates the area designated "at the controls".

442. 2 (13.5.1.3)

The description of administrative procedures should include Fire Protection and Temporary Procedures of an administrative nature.

'0 t

Docket Nos.

50-387 and

-388 Distribution OCT 0 2 1978 NRC PDR Local PDR Docket File LWR 83 File R.

Boyd D. Yassallo F. Williams

0. Parr S. Miner Mr. Norman W. Curtis Vice President - Engineering and Construction D.

Ross Pennsylvania Power and Light Company '

North Ninth Street Allentown, Pennsylvanfa 18101 w/enclosure J. Knight R. Tedesco R.

DeYoung V. Moore R. Vollmer M. Ernst R. Denise ELD IE (3)

(w/extra copies)

BCC:

NSIC TIC ACRS (16)

Dear Hr. Curtis

SUBJECT:

SUS(UEHANNA STEAth ELECTRIC STATION UNITS NOS.

1 AND 2 RE(UEST FOR ADDITIONAL INFORMATION As a result of our review of your applfcatfon for operating licenses for the Susquehanna Steam Electric Plant we find that we need additional information in the area of containment design and testing.

The specffic information requfred is listed in the Enclosure.

The review did not include the informatfon presented in the Desfgn Assessment Report (DAR).

We are currently finalizing our acceptance criteria for the pool dynamic loads and will transmit them to you as soon as they are completed.

Please inform us within 10 days after recefpt of this letter of the date when this requested additional information will be available for our review.

Please contact us if you desire any dfscussion or clarification of the information requested.

Sincerely, original Qs+~ ~

01IIn PRI2 Olan Q. Parr, Chief Light Water Reactors Branch Ho. 3 Division of Pro5ect Management

Enclosure:

As stated cc w/enclosure:

See next page I

l OPPICt~

=I OURNAMt+

DATt+

LHR83:LPM SMiner:ceh.

9/29 /78 L~II I:BC 9/>y/7S I

NRC FORM 518 (9-76) NRCM 0240 4 ul o. oovtRNMiNTPRINTINo OPPlca lofe otaaaR

Hr. Norman W. Curtis cc:

Nr. Earle f4. Head Project Manager Pennsylvania Power

& Light Company 2 North Ninth Street Allentown, Pennsylvania 18101 Jay Silberg, Esq.

Shaw, Pittman, Potts Trowbridge 1800 tl Street, H.

W.

Washington, D. C.

20036 Nr. William E. Barberich, Nuclear Licensing Group Supervisor Pennsylvania Power

& Light Company 2 North Ninth Street Allentown, Pennsylvania 18101 Edward H. Hagel, Esquire General Counsel and Secretary.

Pennsylvania Power

& Light Company 2 North Ninth Street Allentown, Pennsylvania 18101 Bryan Snapp, Esq.

Pennsylvania Power

& Light Company 901 Hamilton Street Allentown, Pennsylvania 18101

ENCLOSURE RE UEST FOR ADDITIONAL INFORMATION SUS UEHANNA STEAM ELECTRIC STATION DOCKET NOS.

50-387 AND 50-388

D21. D D21.22 Containment S stems Branch Based on our review of the information presented in subsection 6.2.1.5 of the

FSAR, we find that the discussion of steam bypass from the dry-well to the wetwell for a small break is incomplete and does not conform to the enclosed branch technical position (BTP) titled "Steam Bypass for the Mark II Containment."

Accordingly, provide the appropriate discussions, justifications, and analyses to demonstrate compliance with the BTP.

021.23 The response to Item 021.05 is inadequate.

Provide a detailed calcu-lation of the friction loss coefficient for the entire vent system.

Discuss whether the results of the 4-T tests have been used to confirm the vent loss coefficient calculated.

State the margin applied to the friction loss coefficient to account for any difference between the Susquehanna vent design and that of the 4-T facility.

021.24 The response to Items 021.6 and 021.7 regarding the vacuum breaker is incomplete with regard to the vacuum breaker set point and opening time and the bases upon which the loss coefficient and vent flow area were determined.

Provide the requested information and in addition:

(1)

Describe the preoperational and inservice tests that will be performed to verify proper pressure setpoint and openi ng time; and

021-2 (2) Provide the sensitivity limits and hysteresis characteristics of the switches.

Provide a discussion and the results of analysis performed to determine the maximum opening between valve disc and seat from when the position indicator system indicates that the valve is closed.

021.25 Section 6.2.1.1.4 of the FSAR states that the containment negative pressure is addressed in the Design Assessment Report.

However, the information as provided is insufficient to allow an independent evaluation.

Therefore, provide the analysis, including the thermo-dynamic model assumptions and the paramters used for the drywell cooldown transients that were performed to establish the contain-ment negative pressure.

021.26 Discuss in detail the design provisions incorporated for periodic inspection and operability testing of the containment heat removal systems'omponents such as pumps, valves, duct pressure-relieving devices and spray nozzles.

021.27 Provide a detailed analysis of the available net positive suction head for the RHR pumps that are used 'as part of the containment heat removal system to demonstrate compliance with Regulatory Gui'de 1.1, "NPSH for Emergency Core Cooling and Containment Heat Removal Systems Pumps."

Specify the required NPSH of the pumps.

021-3 021.28 Describe the sizing analysis performed for the RHR suction screens.

Provide a drawing that shows the suction screen assembly.

021.29 Estimate. for a representative break location, the amount of insu-lation that would be removed from pipes as a result of a LOCA.

On the basis of the properties and characteristtcs of these materials, determine the locations it would accumulate and in what form and whether or not there is a potential for inhibiting suction flow due to clogging of the strainers.

Ogl.30 Provide an analysis of the pr essure and temperature response in the secondary containment due to a postulated LOCA in the primary contain-ment.

Discuss and justify the assumptions made in the analysis and specify the design leakage rate of the reactor building.

021.31 Identify all openings provided for gaining access to the secondary contairment, and discuss the administrative controls that will be exercised over them.

Discuss the instrumentation to be provided to monitor the status of the openings and whether or not the posi-tion indicators and alarms will have readout and alarm capability in the main control room.

021.32 The following additional information related to potential bypass leakage paths is needed.to provide an adequate response to Item 021.03.

(1)

Expand Table 6.2-15 to include any branch lines which penetrate the secondary containment and connect to system lines which penetrate primary containment.

(2)

For each line in (a), identify each of the potential leakage barriers; (3)

For each air or water seal, perform an analysis that will demonstrate that a sufficient inventory of the fluid is available to maintain the seal for 30 days, and describe the testing program and proposed entries for the Technical Specifications that trill verify the assumptions used in the analysis.

Provide the basis for the valve fluid leakage used in the analysis; (4)

For each of these paths where water seals eliminate the poten-tial for bypass

leakage, provide a sketch to show the location of'he water seal relative to system isolation valves; (5)

Table 6.2-15 does indicate that the combustible gas sampling system is eliminated as bypass leakage paths.

Show how this system meets each of the requirements specified in Branch Technical Position, CSB 6-3, Section 9a-f, for a closed system;

021-5 (6)

Table 6.2-15 does not appear to list all potential bypass leakage paths (e.g.,

steam to RCIC system).

Therefore, provide a list of all containment penetrations and seals which do not terminate in the secondary containment and an evaluation of these lines as delineated in the Branch Tech-nical Position, CSB 6-3, "Determination of Bypass Leakage Paths in Dual Containment Plants";

{7)

Table 6.2-22 indicates that the feedwater lines and purge exhaust are secondary containment bypass leakage paths; Table 6.2-15 indicates the opposite.

Discuss the discrepancy.

(8)

The statement is made in Section 6.2.3.2.3 that no bypass leakage will occur following the design Basis LOCA.

Table 6.2-15 identifies those lines penetrating the primary contain-ment which do not terminate inside the secondary containment and are considered as potential bypass leakage paths.

Explain the inconsistency.

021.33 The penetrations taken from Table 6.2-12 and listed below and the corresponding valve arrangement in Figur e 6.2-44 are not consis-tent.

Please clarify these inconsistencies.

The penetrations are:

X 10, 11, 13A. 19, 23, 24, 36, 53, 54, 55, 56, 85 A 8 B, 86 A 5 B, and 215.

021-5 021.24 Section 6.2.4 of the

FSAR, Containment Isolation System,"

should be augmented to provide the justification for any penetration including branch lines which do not conform to the requirements of the General Design Criteria.

In addition provide the containment isolation rationale for your design (e.g.,

RHR pump suction).

021.35 Standard Review Plan 6.2.4, "Containment Isolation Systems,"

states that provisions should be made to allow the operator in the main control room to know when to isolate systems that require remote-manual isolation.

Expand Table 6.2-12 to identify for those systems that rely on remote-manual isolation and the leakage detection provisions for these systems to assure tha adequate information is available to the operator for identifying the affected line and for isolating it.

021.36 (1)

The statement is made in Section 6.2.4.1 of the FSAR that instrumentation lines are designed to the provisions of Regulatory Guide 1.11.

Provide the analysis performed which demonstrates that in the event of a rupture of the instrument lines and/or any component in the line outside the primary containment, the integrity and functional performance of secondary containment and its associated filtration systems are maintained.

021-7 (2)'evise table 6.2-12 to include the isolation provisions for instrumentation lines penetrating the primary containment, D21.37 Revise Table 6.2-12 to include the isolation provisions for the following penetrations X-35B, 36, 37A, 38A, 41, 42, 88 and 93.

D21.38 Table 6.2-12 indicates that the isolation provisions for the con-tainment spray system (X-39A, 200A), the floor drain (X-72A), the equipment drain (X728), the RHR pump suctions (X-203A,C), the RHR pump test line and containment cooling (X-204A), the core spray pump suction

{X-206A), core spray pump test and flush (X-207A), core spray min. recirculation

{X-208A), HPCI pump suction (X-209) RCIC pump suction (X-214),

RHR min. recirc.

(X-226A), suppression pool clean up and drain (X-243) and RHR relief valve discharge (X-246A) conform to the requirement of General Design Criteria 54.

It is our position that the isolation provision to these lines should meet the requirement of GOC 56.

However, a single isolation valve outside containment is acceptable as discussed in Standard Review Plan 6.2.4, II.3.e.

Revise Table 6.2-12 to reflect our position and indicate if the other acceptable alternative for meeting the requirement of the GOC as specified in the SRP could be applied to any of these lines.

021-8 D21.39 Table 6.2-12 indicates that a check valve outside the containment is considered as a contairment isolation valve for the standby liquid control (X-42), the HPCI pump minieum flow recirculation (X-211), the HPCI turbine exhaust (X-210), RCIC pump recirculation (X-217), RCIC vacuum pump discharge (X-245).

Provide justification for this approach.

021.40 The statement is made in Section 6.2.5.2 that nitrogen gas will be used for primary containment atmosphere control.

Discuss the reasons which necessitate inerting the primary containment, since the hydrogen concentration does not exceed 3.5 volume percent.

021.41 The statement is made in Section 6.2.5.3 of the FSAR that the recombiners and purge systems are activated when the hydrogen concentration reaches 3.5 volume percent.

It is our position that the combustible gases resulting from a postulated loss of coolant accident should be controlled without release of radio-active materials to the environment.

Therefore, the hydrogen purge system sh:uld only be used if, as a result of a post-LOCA

event, both recombiner systems fail.

Revise your state to indicate conformance with tnis posit1on.

021.42 Section 6.5.3.1 of the FSAR states that the containment purge system is manually operated from the control room at the discre-tion of the operator.

It is our position that the purge system

021-9 design should satisfy Branch Technical Position, CSB 6-4, "Contain-ment Purging During Normal Plant Operation," if it is used during the reactor operational modes of power operation, startup, hot standby and hot shutdown, or the purge line isolation valve should be locked closed.

Therefore, propose a purge system design that complies with the design provisions of the BTP.

Also, provide the analysis identified in the BTP if the purge system is used during the operational modes specified above.

021.43 The response to guestion 021.14 is incomplete with regard to the requirement imposed on the Reactor Building Ventilation System in order to perform all Appendix J testing.

Provide this information.

021.44 Table 6.2-22 identifies certain valves for which test pressure is not applied in the same direction as the pressure existing when the valve is required to perform its safety function, as required by Appendix J to 10 CFR 50.

Demonstrate that the valve leakage rate is equivalent to or conservative with respect to that which would occur if the test pressure was applied in the direction when the valve is required to perform its safety function.

021.45 Identify those fluid lines penetrating the containment which will be vented and drained to ensure exposure of the system contgfetgnt isolation valves to the containment atmosphere and the ful) differ-ential pressure during the containment integrated leakage rate (Type A) test.

Discuss the design provisions that will permit this to be done.

Those systems that will remain fluid filled for Type A test shoul'd be identified and justification provided.

021.46 Provide plan and elevation drawings of the air locks, and identify all mechanical and electrical penetrations.

Discuss and schemati-cally show the design provisions that will permit airlock door seals and the entire airlock to be tested.

021.47 Discuss the design capability of the door seals to be leak tested at a pressureof Pa; i.e., the peak calculated containment internal pressure.

If it will be necessary to exert a force on the doors to prevent them from being unseated during leak testing, describe the provisions for doing this and discuss whether or not the mechanism can be operated from within the air lock.

Also discuss how the force exerted on the door will be monitored.

021.48

. Discuss your plans for including the reactor building pressure sensing lines, that wi'll become extensions of the containment boundary followirg a LOCA, in Type A test.

021.49 Closed systans outside contaitment having a post accident function, become extensions of the containment boundary following a LOCA.

Certain of these systems may also be identified as one of the redundant containment isolation barriers.

Since these systems may circulate contaminated water or the containment atmosphere.

system components which may leak are relied on to provide contain-ment integrity.

Therefore, discuss your plans for specifying a

leakage limit for each system that becomes an extension of the contairment boundary following a LOCA, and leak testing the sys-tems either hydrostatically or pneumatically.

Also discuss how the leakage will be included in the radiological assessment of the site.

021.50 Table 6.2-22 of the FSAR indicates that excemptions to 10 CFR 50 are required for certain lines.

However, the nature and the rationale for the exemption are not given.

Provide this infor-mation for the following penetrations:

X-10, 21, 23, 24, 85, 86AImB, 87, 93, and 218.

Branch Technical Position Steam B

ass for the Hark II Containment (1)

B ass Ca abilit The Hark II containment should have a steam bypass capability for small breaks of the order of 0.05 ft~ (A/J K).

(a)

Containment Wetwell S ra s

The wetwell spray system including the electrical instrumentation and controls, should meet the standards appropriate to engineered safety features; i.e., quality, redundancy, testability and other appropriate criteria.

The wetwell spray should be automatically actuated ten minutes following a LOCA signal.

In addition, the instrumentation and control systems provided to actuate the wetwell spray should be actuated by diverse parameters.

If the existing wetwell spray system is to be used to improve the bypass capability, the consequences of actuation of the wetwell spray system on ECCS function and long term pool cooling considerations should be evaluated to show that minimum ECCS and pool cooling requirements are met.

(b)

Transient B

ass Ca abilit Anal ses Transient analyses should be provided to establish the capability for a small break.

A normal plant shutdown time of 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> should be assumed.

The results and bases for the analyses

should be provided including the following:

the pressure history in the drywell and the wetwell; identification and quantification of the static heat sinks and the condensing heat transfer coefficient; spray capacity, efficiency, coverage, start time and temperature history; identification and quantification of heat sources.

(2)

Leaka e Tests and Surveillance Re uirements (a)

Hi h Pressure Leak Test' single preoperational high pressure leakage test should be performed on each Mark II unit.

The purpose of this test is to detect leakage in the drywell to suppression chamber vent piping, penetrations, downcomers, vacuum breakers, floor seals, vent seals and the diaphragm.

This test should be performed at approximately the peak drywell to wetwell differential pressure following the high pressure structural test of the diaphragm.

(b)

Low Pressure Leak Tests Post operational low pressure leakage tests should be performed on each Mark II unit to detect leakage in the drywell to suppression chamber vent piping, penetration downcomers, vacuum breakers, floor seals, vent seals and the diaphragm.

This test should be performed at each refueling outing at a differential pressure corresponding to.approximately the submergence of the vents.

%3 a (c)

Acce tance Criteria for Leaka e Tests Thd acceptance criteria for both the high and low pressure leakage tests shall be a measured leakage less than ten percent of the capability of the containment to accommodate bypass leakage at the test pressure.

(d)i Surveillance Re uirements A visual inspection should be conducted to detect possible leak paths at each refueling outage.

Each vacuum relief valve and associated piping should be checked at this time to determine that it is clear of foreign matter.

( 3),

Yacuum Relief Yalve Re uirements (a)

Position Indicators and Alarms P

Redundant position indicators should be placed on all vacuum breakers with indication and redundant alarms in the control room.

The vacuum breaker position indicator system should be designed to provide the plant operators with continuous surveillance of the vacuum breaker position.

The indicators should have adequate sensitivity to detect a total valve opening, for all valves, 'that is less than the bypass capability for a

small break.

The detectable valve opening should be based on the assumption that the valve opening is evenly divided among all the vacuum breakers.

4 (b)

Vacuum Valve 0 erabilit Tests A11 vacuum breakers should be operability tested at mpnthly intervals to assure free movement of the valves.