ML17297A738

Forwards Responses to Questions 281.5-281.8,requested in NRC 810813 Ltr.Also Forwards marked-up FSAR Pages Which Will Be Incorporated Into Future Amend

ML17297A738
Person / Time
Site:	Palo Verde
Issue date:	08/25/1981
From:	Van Brunt E ARIZONA PUBLIC SERVICE CO. (FORMERLY ARIZONA NUCLEAR
To:	Tedesco R Office of Nuclear Reactor Regulation
References
RTR-NUREG-0737, RTR-NUREG-737, TASK-2.B.3, TASK-TM ANPP-18740-JMA, NUDOCS 8109010110
Download: ML17297A738 (49)

Text

REGUL RY INFORMATION DISTRIBUT SYSTEM (RIDS) 1 ACCESSION NBR: 8109010110DOC DATE'! 81/08/25 NOTARIZKD!'ES DO~>>>>

FACIL>>:STN~50 528 Palo Verde Nucl

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cari Statfonr Unft STN-50-529 Palo Verde Nuclear Statfoni Unit 2R Arizona Publi ii Ar izona Publi 05000529 STN-50-530 Palo Verde Nuclear Stations Unit 3i Ar.fzona Publi 05000530 AUTH>>, NAME" AUTHOR AFFILIATION VAN BRU>>4TR Kr, K", i Ar zona Publ f c Service Co.

REC IP ~ >>4AMKr RECIPIENT AFFIL>>IATION TEDESCO RL>>, Assistant Director i for Licensing

SUBJECT:

" For wards responses to Questions" 2815 281,8Rr equested in>> NRCr 810813 ltr ~ Also forwards marked up FSAR pages which will into future'mend be'ncorporated

~

DISTRIBUTION CODEi: B001S COPIES RECEIIVED! L>>TR "

ENCL>> ~ SIZE'::

TITLKi: PSAR/FSAR AMDTS and Related Correspondence NOTES:Standardized Plant. 1 cy1.C Grimes 05000528 Standardized Plant,1 cy!C Grimes 05000529 Standardized Plant F 1 cy:Ci Grimes 05000530 RECIPIENT COPIES RECIPIENT COPIES ID CODE/NAME) LTTR ENCL' ID COOK/NAME LTTR ENCL ACT'ION: A/D LI CEi4SNG L>>I C BR ¹3 BC 1 0 L>>IC" BR ¹3 LA 1 0 KKRRIGANrJ ~ 04 1 1.-

INTERNAL>>; ACCXD E>>VAL 1 1 AUX SYS BR 27 1 1 09 BR26'HEM E>>4G BR 11 1 CONT SYS BR 1 1 CORE PERF BR 10. 1 1 EFF TR SYS BR12 1 1 EMRG PRP DEV 35 1 1 EMRG PRP LIC 36 3 3-EQUIP QUAL BR13>> 3 3 FEMA REP DIV 39 1 1 GEOSCIENCES 28 2 2 HU>>>>l FACT KNG 40 1 1 HYD'/GKO BR 30. 2 2 ILC SYS BR 16 1 1 ILKI 06" 3 3 L>>I C GU ID BR 33 1 1 LIC, QUAL BR 32", 1 MATLI ENG BR 17 1 1 MECH E>>4G BR 18 1 MPA 1 0.

OELD 1 0 OP LZC BR 34 1 1 ~

POINTER SYS BR 19 1 1 PROC/TST REV 20 1 1 QA BR 21 1 1 SS 8R22' 1 1 REAC SYS BR 23>> 1 1 FILE 01 1 1.

SIT'NAL BR 24 1 1 ST ENG BR25 1 1 EXTERNAl.i: ACRS 41 16 16 LPDR 03 1 1 NRC PDR 02~ 1 1 i4S IC 05 1 1 NTIS 1 1

&~ sr TOTAL NUHBER OF COPIED REQUIRED:', LiTTR P1 ENCL'.

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g rmnazm mswvmeaa amzmmmr PHOENIX, ARIZONA 85036 STA. P.o. BOX 21666 A'ugust 25, 1981 ANPP-18740 JMA/KWG Mr. R. L. Tedesco Assistant Director for Licensing Division of Licensing Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Washington, D. C. 20555

Subject:

Palo Verde Nuclear Generating Station (PVNGS) Units 1, 2 and 3 Docket Nos. STN-50-528/529/530 File:. 81-056-026'.1.10

Reference:

Letter from R. L. Tedesco to E. E. Van Brunt, Dated August 13, 1981,

Subject:

Chemical Engineering Branch Questions

Dear Mr. Tedesco:

Please find attached a copy of our responses to the questions 281.5 281.8. Also included are marked up FSAR pages which show this information which will be incorporated in a future amendment.

If you have any questions, please contact me.

Very truly our euu &~~4 E. E. Van Brunt, Jr.

APS Vice President Nuclear Projects ANPP Project Director EEVBJr/KWG/pc Attachment cc: J. Kerrigan (w/a) goo/

P. Hourihan 5 A. C.'Gehr i/(

Bi090iOiiO 8i0825 PDR ADOCK 05000528' IPDR~

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Mr. R. L. Tedesco August 25, 1981 Page Two I, Edwin E. Van Brunt, Jr., represent that I am Vice President Nuclear Projects of Arizona Public Service Company, that the foregoing document has been signed by me on behalf of Arizona Public Service Company with full authority so to do, that'I have read such document and know its contents, and that to the best of my knowledge and belief, the statements made therein are true.

i+,~C +

Edwin "E. Van Brunt, Jr.;;,'~

0 w Sworn to before me thfs~tu+day of 1981.

otary Public My Commission expires:

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FSAR UESTION 6A.62 281.5 In your response to our request for information 281.1, you (6.1.1, 6.5.2) stated that no surveillance of the trisodium phosphate (TSP) baskets will be conducted other than to assure that the baskets are full. It is our position, as stated in CE-PWR Standard Technical Specification (3/4.5.2), that this ECCS subsystem shall be demonstrated operable at least once per 18 months by (1) verifying that a minimum total volume of solid granular TSP dodecahydrate is contained within the TSP baskets, and (2) verifying that when a r'epresentative sample of TSP from a TSP basket is submerged, without agitation, in borated water from the Refueling Water Tank, the pH of the mixed solution is raised to an acceptable level within 4 hours. Indicate that these sur-veillance requirements will be met.

Also, provide the minimum total volume of TSP to be stored in the TSP basket and state the basis for the stored quantity.

RESPONSE

The response is given in the amended response to Question 6A.4.

The minimum total volume of TSP is 154 ft (12 tons anhydrous trisodium phosphate). This minimum volume was calculated by CE to result in a "worst case" minimum sump solution pH of approximately 7.0. PVNGS has installed (in each unit) 9 baskets of dimensions 2' 2' 2'nd 8 baskets of 4' 4' 4'.

The available volume of these baskets is 684 ft .

PVNGS FSAR APPENDIX 6A sodium hydroxide solutions to preclude iodine evolution do not apply to PVNGS.. Initial spray pH levels are about 5.3, and within two hours of spray initiation, sump pH levels are stipulated to be maintained between 7.0 and 8.5. (See sections 6.1.1 and 6.5.2.) Even with sump pH of less than 7.0, iodine does not evolve from the sump.

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Indicate the total amount of protective coatings, paints, and organic materials (including uncovered cable insulation) used inside the containment that do not meet ANSI N101.2 (1972) and Regulatory Guide 1.54.

RESPONSE: Within the containment of Unit 1, there are cur-rently less than 4000 ft2 (of a total of about 393,600 ft )

of coated surfaces documented not to conform with ANSI N101.2 and Regulatory Guide 1.54. Exceptions taken to Regulatory Guide 1.54 are given in section 1.8 and include:

~ About 4,000 sq. ft. of insulated piping

~ About 140,480 sq. ft. enclosed within cabinets or enclosures.

Q" *' "! N" Q"" "* (1.8 and 6.5)

In Section 1.8 of the FSAR which deals with 'Conformance to NRC Regulatory Guides', reference is made to Regulatory Guide 1.52, Rev. '0 (June"1973 ) and Rev. 1 (July 1976) versions. Since Regulatory Guide 1.52, Revision 2, (March 1978), "Design, Testing and Maintenance Criteria for Post-Accident Engineered-Safety-Feature Atmosphere Cleanup System Air Filtration and Adsorption Units of Light-Water-Cooled Nuclear Power Plants" Amendment 4 6A-2 May 1981

1 I A

Insert A to Pa e 6A-2 The. following surveillance of the trisodium phosphate baskets shall be conducted at least once per 18 months:

a) Verifying that a minimum total of g),ggr,'ubic feet of solid gran-ular trisodium phosphate dodecahydrate (TSP) is contained within TSP storage baskets.

b) Verifying that when a representative sample of TSP from a TSP storage basket is submerged, without agitation, in borated water from the Refueling Water Tank (RWT), the pH of the mixed solution is raised to greater than or equal to 6.0 within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

FSAR Question 6A.63 281.6 In Table 6 ~ 1-4 of the FSAR, you indicate that there are 259,560 pounds

/ il', 'l of cable insulation inside the containment building. Indicate what fraction of thi's weight consists of organic materials. We also need the following additional information for estimating the generation rate of combustible gases from organic materials in cable insulation under DBA conditions: (1) The quantity (weight and volume) of uncovered cable and cable in closed metal conduit or closed cable trays'e will give credit for beta radiation shielding for that portion of cable that is indicated to be in closed conduit or trays, (2) A breakdown of cable diameters and associated conductor cross section, or an equivalent cable diameter and conductor cross section that is representative of total cable surface area associated with the quantity of cable identified in 1) above. (3) The major organic polymer or plastic'aterial in the cables. If this information is not provided, we will assume the cable insulation to be polyethylene and assume a G value for combustible gas of 3.

Response

Table 6.1-4 has been revised in Amendment 5 to indicate 39,500 (37600 + 5%)

pounds of cable insulation within the containment.

(1) The quantity of covered/uncovered cable is not readily available.

Therefore, all cable is assumed exposed.

(2) 'A breakdown of cable diameters and conductor cross section is attached as Table 281.6-1.

(3) The major organic polymer in the cable insulation is flame-retardant ethylene propylene rubber and chlorinated polyethylene. (Other organic polymers in various insulation types are listed in Table 6.1-4.)

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Insert A to Pa e 9A-'22 The fuel pool cleanup filter cartridges will be replaced when the pressure drop across the filters exceeds 25 psid during operation of the system, or the gamma radiation reading on the outside of the filter housing exceeds a reading to be determined, when normal system radiation levels have been measured in order to maintain radiation exposures ALARA.

Fuel pool cleanup ion exchanger performance will be monitored by determining a decontamination factor from ion exchange inlet to outlet samples for gross activity and gross iodine. Ion exchanger resin will be replaced when the decontamination factor drops below a minimum efficiency level, as determined by examining initial operating decontamination factors, for 3 consecutive samples taken 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> apart or when the gamma radiation readings on the outside of the filter housing exceeds a reading to be determined when normal system radiation levels have been measured in order to maintain radiation exposures ALKQ..

Your response to our previous Question 281.2 did not indicate any chemical or radionuclide limits for initiating replacement of filters and ion exchange resins. It is our position that chemical and radionuclide limits in the spent fuel pool water, such as conductivity, gross gamma and iodine activity, demineralizer differential pressure, pH,.and crude level, are needed for initiating corrective action to enable safe operating conditions in the pool. Verify that you will meet this position and provide the above information.

RESPONSE

The response is given in the amended response to Question 9A.34.

. PVNGS FSAR APPENDIX 9A Q"" (9.1.3)

For the fuel pool cleanup system, indicate that chemical analyses at least weekly and continuous radiological monitor-ing will be made for measuring the efficiency of the filters and ion exchange resins to remove impurities and radioactive materials from the pool water. State what criteria (chemical parameters, decontamination factors, etc.) will be used to determine replacement of the filters and ion exchange resins.

RESPONSE: For the fuel pool cleanup system, chemical analysis down stream of the filters and ion exchangers is done via batch sampling only. These samples as a minimum are to be taken at one week intervals.

Continuous radiological in-line monitoring is not a

. requirement for the system as outlined in Regulatory Guide 1.45 (also, see section 11.5.1 and table 9.3-3).

High electrical conductivity or radiation levels in the batch sample are the criteria used to determine replacement of the resin bed for the ion exchangers.

The cleanup filters are provided with pressure differential indicator switches which give local indication of pressure drop across the filters and alarm to the lo~cal anel on high pressure drop.',,!, - >', . r i op Jw~.'i =,!",..' '41.b CndViging C lr4r3.f, )<","-

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Describe the provisions to meet the requirements of post-accident sampling of the primary coolant and containment atmosphere. The description should address all the require-ments outlined in Section II.B.3 of Enclo'sure 3 in NUREG-0737 Amendment 4 9A-22 Nay 1981

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FSAR uestion 9A.60 281.8 Provide information that satisfies the attached proposed license (9.3.2, II.B.3) conditions for post-accident sampling.

NUREG-0737, II.B. 3 Post Accident Sam lin Ca abilit Provide a capability to obtain and quantitatively analyze reactor coolant and containment atmosphere samples, without radiation exposure to any individual exceeding 5 rem to the whole body or 75 rem to the extremities (GDC-19) during and following an accident in which there is core degradation.

Materials to be analyzed and quantified include certain radionuclides that are indicators of severity of core damage (e.g., noble gases, iodines, cesiums and non-volatile isotopes), hydrogen in the containment atmosphere and total dissolved gases or hydrogen, boron and chloride in reactor coolant samples in accordance with the requirements of NUREG-0737.

To satisfy the requirements, the applica nt .'hould (1) review and modify his sampling, chemical analysis and radionuclide determination capabilities as necessary to comply with NUREG-0737, II.B.3, (2) provide the staff with information pertaining to system design, analytical capabilities and procedures in sufficient detail to demonstrate that the requirements have been met.

EVALUATION AND FINDINGS The applicant has committed to a post-accident sampling system that meets the requirements of NUREG-0737, Item II.B.3 in Amendment , but has not provided, the technical information required by NUREG-0737 for our evaluation.

Implementation of the requirement is not necessary prior to low power operation because only small quantities of radionuclide inventory will exist in the reactor coolant system and, therefore, will not affect the health and safety of the public. Prior to exceeding 5% power operation, the applicant must demonstrate the capability to promptly obtain reactor coolant samples in the event of an accident in which there is core damage consistent with the conditions stated below:

1. Demonstrate compliance with all requirements of NUREG-0737, II.B.3, for sampling, chemical and radionuclide analysis capability, under accident conditions.

2. Provide sufficient shielding to meet the requirements of GDC-19, assuming Reg. Guide 1.4 source terms.

3. Commit to meet the sampling and analysis requirements of Reg. Guide 1.97, Rev. 2.

4. Verify that all electrically powered components associated with post-accident sampling are capable of being supplied with power and operated, within thirty minutes of an accident in which there is core degradation, assuming loss of offsite power.

5. Verify that valves which are not accessible for repair after an accident are environmentally qualified for the conditions in which they must operate.

6. Provide a procedure for relating radionuclide gaseous and ionic species to estimated core damage.

7. State the design or operational provisions to prevent high pressure carrier gas from entering the reactor coolant system from on line gas analysis equipment, if it is used.

8. Provide a method for verifying that reactor coolant dissolved oxygen is at C O.l ppm if reactor coolant chlorides are determined to be 0.15 ppm.

9. Provide information on (a) testing frequency and type of testing to ensure long term operability of the post-accident sampling system, and (b) operator training requirements for post-accident sampling.

In addition to the above licensing conditions, the staff is conducting a generic review of accuracy and sensitivity for analytical procedures and on-line instrumentation to be used for post-accident analysis. We will require that the applicant submit data supporting the applicability of each selected analytical chemistry procedure or on-line instrument along with documentation, demonstrating compliance with the licensing conditions four, months prior to exceeding 5% power operation, but review and approval of these procedures will not be a condition for full power operation. In the event our generic review determines a specific procedure is unacceptable, we will require the applicant to make modifications as determined by our generic review.

RESPONSE

The PVNGS design uses in-line chemical and radionuclide analyses with backup grab sample capability to meet the requirements of NUREG-0737. These analyses fully meet the ranges and accuracies specified in Reg. Guide 1.97, Rev. 2.

The capability exists to sample'nd analyze the required samples over the required ranges within the allotted three hour time span. Table 1 provides further information on sample points and analyses. All samples, including grab samples, may be taken without exposing any individual to radiation levels in excess of GDC 19 criteria. Isotopic analysis is provided to identify and

quantify the isotopes of the nuclide categories corresponding to the source terms given in Reg. Guide 1.4 and 1.7. A manually loaded Class IE power source is'vailable within 30 minutes of an accident that assumes loss of offsite power. All equipment in the post-accident sampling system (PASS),

including valves, piping, cabling, motors and analyzer electronics, is qualified to the environment it would experience under accident conditions.

The PASS computer will determine estimated core damage by comparing the fraction of the core inventory in the RCS against the known core isotopic inventory. All listed analyses can be performed on any sample taken. This provides a complete picture of radiological and chemical data for historical comparison. Carrier gases are not used in the analyses.

The entire PASS system will be operated from a central computer which directs a controller to sample through valve and pump operation sequences. All data will be analyzed, stored and displayed on the computer. Ma-ual control will 1 be available on loss of the automatic functions. Grab samples can be taken remotely during post-accident conditions to minimize operator'exposure.

During normal operations, grab samples will be taken locally at the analyzer module.

Dissolved oxygen is verified to be below 0.1 ppm by the dissolved oxygen analyzer.

The PASS will provide weekly data during normal operations to meet various chemical and radiological technical specification sample times. Grab samples

will be taken on a routine basis to assure technician familiarity. Testing frequency will be determined by manufacturers requirements to maintain reliability. Operator training requirements will be determined at a later date.

Refer to Section 9.3.2 and PVNGS LLIR Section II.B.3 for further details of the PASS.

Refer to PVNGS LLIR Section II.B.2 for further details on plant shielding.

TABLE 1 FOR NRC QUESTION 281.8 (9.32 and II.B.3)

Post Accident Sampling System

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~Lt u1d Containment Recirculation Sump, Reactor Coolant System, Containment Radwaste Sump, Safety Injection A, Safety Injection B, Auxiliary Building Sumps, Letdown System Gas Containment Air, Volume Control Tank

~Out uts

~LL udd Equipment Drain Tank, Reactor Drain Tank Gas Containment Air

~Anal ses

~LX uld Gamma Spectrum 10 uCi/ml to 10 Ci/ml Isotopic Gross Gamma 10 uCi/ml to 10 Ci/ml Gross Boron 0 to 6000 ppm pH 1 to 13 (temperature compensated)

Dissolved Oxygen 0 to 20 ppm Dissolved Hydrogen 0 to 2000 cc (STP)/Kg Chlorida 0 to 20 ppm Gas

-3 -1 Gamma Spectrum 10 uCi/ml to 10 Ci/ml Isotopic Gross Gamma 10 uCi/ml to 10 Ci/ml Gross Gaseous Oxygen 0 to 30%

Gaseous Hydrogen 0 to 10% (provided by the Containment Hydrogen Control System)

Chemical analyses are done in parallel. Chemical analyses are performed while the radiological analyses are being performed.

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