ML20205S027

Forwards Rept of Changes Made During 1985 Per 10CFR50.59

ML20205S027
Person / Time
Site:	Palo Verde
Issue date:	05/21/1986
From:	Van Brunt E ARIZONA PUBLIC SERVICE CO. (FORMERLY ARIZONA NUCLEAR
To:	Martin J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION V)
References
ANPP-36659-EEVB, NUDOCS 8606060138
Download: ML20205S027 (32)

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Arizona Nuclear Power Project P O. box 52034 e PHOENIX, ARIZONA 850724034 May 21, 1986 ANPP-36659-EEVB/BJA/98.05 Mr. John B. Martin, Regional Administrator Office of Inspection and Enforcement U.S. Nuclear Regulatory Commission Region V 1450 Maria Lane, Suite 210 Walnut Creek, CA 94596-5368

Subject:

Palo Verde Nuclear Generating Station (PVNCS)

Units 1, 2 and 3 Docket Nos. : STN 50-528 (License No. NPF-41)

STN 50-529 (License No. NPF-51)

STN 50-530 Annual Report of Changes Made Pursuant to 10 CFR 50.59 File: 86-056-026

Dear Mr. Martin:

At tached please find a report of those changes at PVNGS which were made pursuant to the requirements of 10 CFR 50.59 during the 1985 calendar year.

The attached report contains a brief description of such changes, tests, and experiments, including a summary of the safety evaluation of each change. By copy of this letter, we are also furnishing a copy of this report to the Director of Inspection and Enforcement.

If you have any additional questions on this matter, please contact Mr. W. F. Quinn of my staff. ,

Very truly yours, f .

LT.\mh m E. E. Van Brunt, Jr.

Executive Vice President Project Director EEVB/BJA/ dim Attachment cc: E. A. Licitra (all w/a)

R. P. Zimmerman A. C. Gehr  !

J. M. Taylor (Director-Office of Inupection and Enforcement) l 860606013e e60521  !

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f *$ Y This a ttachment contains a listing of the changes, tests, and experiments which were made pursuant to the requirements of 10 CFR 50.59 during the 1985 calendar year at Palo Verde Nuclear Generating Station. A brief description of the change and a summary of the safety evaluation for each change is presented below. It should be noted that none of the changes involved an unreviewed safety question. Specifically, none of these changes: (1) increased the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in tha safety analysis report; or (ii) created the possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report; or (iii) reduced the margin of safety as defined in the basis for any technical specification. Additionally, some of these changes were previously submitted to the NRC staff for review prior to the receipt of a low-power Operating License for PVNGS Unit 2. These changes are noted along with the date of the previous submittal.

(1) Description of Change Revised the examination criterion for resuming licensed duties following the annual evaluation examination for licensed operators. The revised criterion stated that a score of at least 70% on each section retaken and an overall average of at least 80% for all sections retaken is required prior to resuming licensed duties.

Summary of Safety Evaluation This change to the annual evaluation examination criterion did not result in an unreviewed safety question. The change was in accordance with the NRC requirements for licensed operator requalification examinations. These requirements are identified in Enclosure 1 of the March 28, 1980 letter from H. R. Denton, NRC, to all power reactor applicants and licensees. The requirement states that the new passing grade for requalification shall be 80% overall and 70% for each category.

(2) Description of Change Revised the safety analysis report to note that the flanges are removed from the 10-inch drain lines from the refueling pool during normal power operations. This precludes the trapping of water in the pools.

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1 Summary of Safety Evaluation This change to identify the fact that the flanges are removed during normal operations did not involve an unreviewed safety question. The flanges are removed from the drain lines during normal operations to ensure that water does not get trapped in the drain lines or in the refueling pool following an accident where sump recirculation is required. With the flanges removed, there is a path for the water to flow down from the refueling pool to the recirculation sumps. This assures that the NPSH requirements for the ECCS pumps will be met. The removal of the flanges is administrative 1y controlled at PVNGS. This change was previously submitted to the NRC in a letter dated August 29, 1985 (ANPP-33291) .

1 (3) Description of Change Inclusion of PVNGS Units 2 and 3 reactor coolant pressure boundary fracture toughness data in the safety analysis report.

Summary of Safety Evaluation Fracture toughness data for reactor coolant pressure boundary components is required to be included in the safety analysis report. This is in accordance with the recommendatiens of Regulatory Guide 1.70 for the standard format and content of safety analysis repo rt s . Thus, this change was made to implement an NRC requirement for information contained in safety analysis reports and did not involve an unreviewed safety qttestion. Additionally, this change was previously submitted for NRC staff review in Amendment 14 of the PVNGS Final Safety Analysis l Report (ESAR) which was submitted on February 28, 1985. l (4) Description of Change Revised the testing duration of ventilation system air filtration units to note that the units will be operated at least once every 31 days for at least 15 minutes. This is a deviation from the recommendations of Regulatory Guide 1.52, Revision 1.

Summary of Safety Evaluation This change did not result in an unreviewed safety question. Regulatory Guide 1.52 recommends operating the atmosphere cleanup units for at least 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> per month. The purpose of this recommendation is to reduce the buildup of moisture on the adsorbers and HEPA filters. Due to the low humidity in the desert area of PVNGS , a 15 minute operating time every month will assure the operability of the filtration units and the moisture buildup will not be a problem in this environment.

Additionally, this change is in accordance with the surveillance require-ents of the PVNGS Technical Specifications which require a mininum 15 minute operating time every 31 days. This change has also been submitted to the NRC staff for review by letter dated August 30, 1985 (ANPP-33312).

(5) Description of Change Modification to the non-licensed operator training program to add formal classroom lectures on plant systems for those individuals without previous experience in nuclear power.

Summary of Safety Evaluation This change will result in a more comprehensive training program for non-licensed operators (auxiliary operators). The change allows for the operators who do not have previous nuclear power experience to receive formal classroom training on plant systems. This will result in better training for non-licensed operators and will allow them to operate equipment more efficiently. Therefore, this change is an administrative change to the training program und is not an unreviewed safety question. It should be noted that this change has been previously submitted to the NRC staff by letter dated August 30, 1985 (ANPP-33314).

(6) Description of Change The FSAR was revised to note an exception from the recommendations of Regulatory Guide 1.16, Revision 1 regarding non-routine reporting requirements.

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Summary of Safety Evaluation Section C.2 of Regulatory Guide 1.16, Revision 1 specified the requirements for non-routine reporting. Subsequent to the issuance of this Regulatory Guide, the NRC has issued revised regulations (10 CFR 50.72 and 10 CFR 50.73) which provide some of the requirements for non-routine reporting. Additional reporting requirements are provided in the PVNGS Technical Specifications. 'Ihu s , these non-routine reporting requirements supercede those contained in Regulatory Guide 1.16, Revision 1. This change was administrative in nature to note the exception to Regulatory Guide 1.16 and does not involve an unreviewed safety question. This change has been previously submitted to the NRC staff by letter dated August 30,1985 (ANPP-33312).

(7) Description of Change Revision to FSAR Section 1.1.5 to note the current scheduled completion dates and commercial operation dates for the three PVNGS units.

Summary of Safety Evaluation This change to the Safety Analysis Report is administrative in nature and does not involve an unreviewed safety question. The change is only required for the FSAR to reflect the current schedules for completion of the units and for expected commercial operation dates for the units.

This change does not modify any plant equipment and has no effect on plant safety. Additionally, this change has been previously submitted for NRC staff review by letter dated August 29,1985 ( ANPP-33289).

(8) Description of Change The preoperational test method for the Boric Acid Batching Tank (BABT) subsystem test was revised due to the fact that the tank is not capable of being drained to the Equipment Drain Tank (EDT). Thus, the required boric acid concentration samples will be obtained while transferring the contents of the BABT to the Refueling Water Tank (RWT) or while the BABT is being drained to the non-ESF sump.

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Summary of Safety Evaluation The preoperational test requirements for the BABT subsystem are presented in CESSAR Section 14.2.12.1.12. The test method described in CESSAR requires boric acid concentration samples to be obtained as the .

! BABT is being drained to the EDT. This test methodology does not reflect the as-built condition of the plant as the BABT drains to the non-ESF sump instead of to the EDT. Therefore, the CESSAR test method ,

} was changed to allow for obtaining boric acid concentration samples as the BABT was being transferred to the RWT or while draining to the

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j non-ESF sump. This change had no effect on the safe operation of'the 1  :

i plant as the required boric acid concentration data is obtained to i I

satisfy the preoperational testing requirement. Therefore, this change ,

j to the test methodology did not involve an unreviewed safety question. l This change has been submitted to the NRC staff by letter dated August -

1 29,1985 (ANPP-33289).

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(9) Description of Change This change eliminates the requirement to . initially test the gas

stripper pumps to demonstrate head and capacity.

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j' Summary of Safety Evaluation i CESSAR Section 9.3.4.4 requires all pumps in the Chemical and Volume Control System (CVCS) to be initially tested to demonstrate head and i capacity. Contrary to this CESSAR requirement, the gas stripper pumps were not tested to demonstrate head and capacity during the j

l preoperational test program due to the fact that there are no available i test connections for measuring suction and discharge conditions. This testing exception has been determined to be acceptable for the following 1

two reasons. Firstly, the gas stripper pumps . are not required to

! operate following any design. basis accidents. Thus, these pumps are not l l needed to mitigate any accidents or to safely shutdown the reactor. l 4

Secondly, these pumps are supplied as part of the gas stripper package and this package was tested and qualified by the vendor. Therefore, the.

deletion of this testing requirement does not compromise. the safety of the plant and does not constitute an unreviewed safety question.

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l This change has been previously submitted to the NRC by letter dated August 29, 1985 (ANPP-33289). 1 (10) Description of Change l A change to a procedure as described in the FSAR to delete the requirement to collect and analyze a sample of the -containment atmosphere immediately after each containment purge batch release.

. Summary of Safety Evaluation A gas sample of the containment atmosphere is required per the PVNGS Technical Specification prior to the initiation of a containment purge

] batch release. The purge is continuously monitored by area radiation j monitors and by the plant vent effluent radiation monitor during the containment purge. Additionally, the containment purge batch release contribution to the plant release is incorporated into the continuous

sample collected at the plant vent release point. The deletion of the i

post purge containment atmosphere sample will have no effect on plant

] safety and does- not result in an unreviewed safety question. This j

change has been previously submitted to the NRC by letter dated August 30,1985 (ANPP-33310).

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i (11) Description of Change l Additional non-metallic insulation was 'added to piping inside

{ containment to reduce heat losses. The FSAR was revised to reflect the i- addition of this insulation.

1 Summary of Safety Evaluation This addition of non-metallic insulation inside containment was i

evaluated for the potential of blocking the recirculation sump screens or damaging the ECCS pump seals and bearings. The results of the analysis are within the NRC acceptance criteria and ensure the j availability of the ECCS pumps following a major accident. Therefore, ,

j this change did not result in an unreviewed safety question. This-change has been previously submitted to the NRC by letter dated August 29, 1985 (ANPP-33291).

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1 l (12) Description of Change I

Changed the FSAR to identify the ANPP Nuclear Engineering Department as 4

the responsible organization for the classification of safety related l items. The previous responsible organization was the PVNGS Engineering

! Department.

Summary of Safety Evaluation l

This change is an administrative change and only involves a change in the departmental responsibility for a classification procedure. This change does not involve a change to any equipment important to safety and does not constitute an unreviewed safety question. Additionally, this change has been previously submitted for NRC staff review by letter 1

i dated August 30, 1985 ( ANPP-33311) .

(13) Description of Change The FSAR was changed to reflect the as-built condition of the plant for l the seismic monitoring instrumentation. Specifically, there is a single I trigger switch to activate the seismic recording system. This trigger

, switch continuously monitors the output level of a single i accelerometer. It should be noted that the previous FSAR description i

implied the use of multiple trigger switches.

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, Summary of Safety Evaluation The function of this single trigger switch is to activate the seismic recording system to record a seismic event. The recorded data is then used to analyze the event after occurrence. Thus, the trigger switch

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and tape recorder have no accident prevention or mitigatiots function and will not affect the safety of the plant. Therefore, this change does i

not result in an unreviewed safety question. This change has also been submitted to the NRC by letter dated August 30, 1985 (ANPP-33312).

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(14) Description of Change The FSAR was revised to reflect the as-built condition of the plant.

Specifically, the combustible loading for fire zone 54 was revised and the equivalent fire severity for this zone was increased to 66 minutes based on this combustible loading.

Summary of Safety Evaluation The increased equivalent fire severity for this zone required a re-evaluation of the fire consequences in this zone. It was determined that no automatic suppression capability is required for this zone due to the existing fire detection instrumentation, the zone perimeter wall, and the manual fire fighting capability. This change did not result in an unreviewed safety question and this change was submitted for NRC staff review by letter dated April 15, 1985 (ANPP-32398).

(15) Description of Change The FSAR was revised as a result of a physical modification which provided for forced air circulation in the pressurizer compartment. Due to the increased subcompartment pressure caused by this modification, the pressurizer subcompartment design wall loading was increased from 54 psid to 73 psid.

Summary of Safety Evaluation The modification to the pressurizer subcompartment decreased the vent area available for relieving steam and water to the containment building in the event of a pipe rupture inside the subcompartment. The pressurizer subcompartment was- re-analyzed to determine if the subcompartment was capable of sustaining the increased peak subcompartment pressure resulting from this modification. The re-analysis showed that the subcompartment is capable of withstanding at least 73 paid. This is greater than the calculated peak pressure in the subcompartment. Therefore , this change did not result in an unreviewed safety question. This change has been previously submitted for NRC staff review by letter dated April 25, 1985 ( ANPP-32497) .

(16) Description of Change The FSAR was modified to indicate that a breathing air compressor is not yet available at PVNGS.

Summary of Safety Evaluation i The breathing air compressor was originally intended to be a backup to the normal air supply. The normal air supply consists of self-contained breathing units with a one hour air supply and a six hour supply of reserve air to replenish exhausted air supply bottles. The self-contained breathing units with the six hour backup air supply are sufficient to meet the requirements of Branch Technical Position CMEB 9.5-1. Therefore, the elimination of the backup breathing air compressor was not an unreviewed safety question. This change has been submitted to the NRC staff by letter dated April 29,1985 (ANPP-32515) .

1 (17) Description of Change Installed ionization smoke detector systems above the false ceiling of the hot and cold labs in the auxiliary building and the computer room in the control building.

Summary of Safety Evaluation The installation of these ionization smoke detectors will improve plant safety by providing an early detection of smoke from a fire which could potentially damage safe shutdown circuits installed in conduit above these false ceilings. These new detectors will not have any adverse impact on other plant systems. Therefore, the installation of the detectors is not an unreviewed safety question. This change has been previously submitted to the NRC by letter dated August 30, 1985 (ANPP-33310).

(18) Description of Change The FSAR was changed to require the control room to be pressurized to 1/8 inches water gauge pressure by the essential ventilation system following a radiological accident. The FSAR had previously required 1/4 inch pressurization.

Summary of Safety Evaluation The control room is maintained at a positive pressure with respect to atmospheric pressure during certain post-accident situations for the purpose of ensuring that the control room operators are not exposed to the harsh environment during an accident. A pressurization of 1/8 inches w.g. pressure is acceptable to perform this function. As long as the control room envelope is maintained at this positive pressure with respect to the surroundings, there will be no infiltration into the control room and the operators will be sufficiently protected.

Therefore, the consequences of a radiation accident will not be increased as a result of this change. This change also meets the guidance of the Standard Review Plan and Regulatory Guides 1.78, 1.52, and 1.95. This change is also in accordance with the requirements of

, PVNGS Technical Specification 3.7.7 which requires the essential ventilation system to be capable of maintaining at least 1/8 inch water gauge pressure with respect to the adjacent areas. There fore , this change does not involve an unreviewed safety question. This change has been submitted to the NRC by letter dated August 30,1985 (ANPP-33291).

(19) Description of Change The FSAR was revised to reflect the as-built condition of the plant.

Specifically, emergency lighting with an eight hour capacity is installed in all areas of the plant where safe shutdown equipment is operated and in access and egress routes to these areas. Emergency lighting with a 1-1/2 hour capacity is installed for personnel egress from areas other than safe shutdown areas.

Summary of Safety Evaluation 4

Emergency lighting with an eight hour capacity is installed in all areas of the plant required for safe shutdown. Thus, the capability to bring the plant to a safe condition after the accident is ensured. This change is in accordance with the applicable regulatory requirements and is not an unreviewed safety question. This change has been submitted to the NRC by letter dated August 30,1985 (ANPP-33306).

(20) Description of Change The FSAR was updated to give the correct l'nstrument identification tag number for the plant vent radiation monitors. The previously listed tag number on FSAR page 9A-25 was not correct.

Summary of Safety Evaluation This change does not involve an unreviewed safety question. This change is administrative and does not involve any physical changes to the plant. Additionally, this change has been submitted to the NRC by letter dated August 30, 1985 (ANPP-33312).

(21) Description of Change The FSAR was revised so that it would be in agreement with previous ANPP commitments that were made in response to NUREG-0737 item I.A.2.3 i

regarding administration of training programs. The specific change involves revising the FSAR to require training instructors who teach systems, integrated responses, transients, or s11ulator courses, to hold or have held an NRC license or certification at the SRO level on a PWR.

Summary of Safety Evaluation This change is administrative in nature and involves revising the FSAR to reflect commitments which were made in a separate document (i.e., the PVNGS Lessons Learned Implementation Report) . This change is in agreement with the NRC requirements which are presented in the March 28, 1980 letter from H. R. Denton, NRC, to all licensees. Therefore, this I

change does not involve an unreviewed safety question. This change has also been submitted to the NRC by letter dated August 30, 1985 (ANPP-33314).

(22) Description of Change Change to utilize a different pressure sensor to initiate the low suction pressure trip of the waste gas compressors of the gaseous radwaste system. The suction pressure trip setpoint for the new sensor is 1.5 psig as compared to a setpoint of 0.5 psig for the original pressure sensor.

Summary of Safety Evaluation The low suction pressure trip is designed to ensure that the waste gas compressors do not operate with a low suction pressure which could potentially damage the equipment. The pressure sensor that is used to initiate the low pressure signal for the compressor suction is changed.

The pressure sensor that was originally used for this purpose was installed as part of the compressor assembly. Due to equipment problems with this sensor, the pressure instrument on the waste gas surge tank was selected to perform this function. A presssure of 1.5 psig at the waste gas surge tank pressure sensor corresponds to a compressor suction pressure of 0.5 psig due to the pressure drop in the lines. Therefore, the low suction pressure setpoint was revised to compensate for line losses. This change did not constitute an unreviewed safety question as the change did not compromise the low suction pressure protection which is provided for the waste gas compressors. This change has been submitted to the NRC by letter dated August 30,1985 (ANPP-33313).

(23) Description of Change Added clarification to the FSAR to define the acceptable Halon concentrations. A maximum Halon concentration of 7% by volume is acceptable for occupied hazard areas and a maximum concentration of 10%

by volume is acceptable in unoccupied areas or in areas that are capable of being evacuated within one minute.

Summary of Safety Evaluation The purpose of this change was to clarify the acceptance criteria for maximum Halon concentrations in different areas of the plant. The acceptance criteria of a maximum of 7% by volume for occupied hazard areas and a maximum of 10% by volume in unoccupied areas or areas easily capable of being evacuated, is in accordance with the acceptance criteria of NFPA Pamphlet 12A (1984). Therefore, this change does not constitute an unreviewed safety question. This change has been submitted for NRC staff review by letter dated August 30, 1985

( ANPP-33310) .

(24) Description of Change Modification to the Quality Assurance program to delete chemical analysis for the list of special processes which are subject to the requirements of 10 CFR 50, Appendix B. Additionally, the special process of cleaning has been clarified as chemical cleaning.

Summary of Safety Evaluation 10 CFR 50, Appendix B requires that measures be established to assure that special processes are controlled and accomplished by qualified personnel using qualified procedures in accordance with applicable codes, standa rds , specifications, criteria, and other special requirements. Due to the programs and procedures utilized at PVNGS, chemical analysis does not fall under the category of special processes. Additionally, cleaning is not considered to be a special process, but chemical cleaning is considered a special process. These changes to the list of special processes are in accordance with the applicable regulatory guidance (ANSI 18.7, 10 CFR 50, App. B, NUREG-0800). Additionally, this change does not result in a decrease of the controls on chemical analysis. Programs and procedures concerning chemical analysis are maintained at PVNGS. This change does not involve an unreviewed safety question. This change has been submitted to the NRC, in accordance with the requirements of 10 CFR 50.55(f), by letter dated September 30, 1985 (ANPP-33603).

(25) Description of Change Changed the preoperational test program to delete the requirement to verify the undervoltage relaying of the Class IE 480V Motor Control Centers (MCCs).

Summary of Safety Evaluation The preoperational test method for the Class IE 480V McCs is presented in Section 14B.7 of the FSAR. The original test method requires the verification of the proper operation of the undervoltage relaying for the 480V MCCs. This test method is not appropriate because the 480V MCCs do not have undervoltage protection. The undervoltage protection

i for the 480V distribution system is at the 480V load center buses and is preoperationally tested in accordance with the test method described in FSAR Section 14B.6. Therefore, this change is required for the FSAR to reflect the as-built condition of the plant and does not involve an unreviewed safety question. This change has also been submitted to the NRC by letter dated August 30, 1985 (ANPP-33309).

(26) Description of Change Modification to the plant to upgrade the doors, walls, and penetration seals for walls surrounding the Control Element Drive Mechanism (CEDM) room. The walls surrounding the CEDM room at the 120 foot level of the auxiliary building have been upgraded to be a 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> fire rated barrier.

Summary of Safety Evaluation This modification to the walJs surrounding the CEDM room enhances the fire zone separation and provides improved protection for the equipment located in the CEDM room. This change results in an increase to the safety of the plant and does not involve an unreviewed safety question.

This change has been submitted to the NRC by letter dated August 30, 1985 (ANPP-33310).

(27) Description of Change A portion of the description of the Quality Assurance (QA) program in Section 17 of the PVNGS FSAR was revised to provide clarification of the intent of the section. Spe cifically, the responsibilities of the ANPP Technical Services Department concerning document control were clarified.

Summary of Safety Evaluation This change to the program description is administrative and only provides clarification to the original program description. Therefore, this change does not involve an unreviewed safety question. This change has been submitted for NRC review by letters dated June 13, 1985 (ANPP-32821) and August 30,1985 (ANPP-33311). NRC-Region V review and t acceptance of the changes is documented in a letter dated July 9,1985 from D. F. Kirsch, NRC, to E. E. Van Brunt, Jr. , ANPP.

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i (28) Description of Change A change was made to the preoperational test description for the l emergency lighting system. The change involved adding more detail to i the acceptance criteria.

i Summary of Safety Evaluation 4

The preoperationel test of the emergency lighting system is described in Section 14B.10 of the PVNGS FSAR. The changes that were made to the test acceptance criteria were for clarification purposes and are in j accordance with the previously approved system design. Therefore, this

change does not constitute an unreviewed safety question as the emergency lighting system is not being changed and the design is thoroughly tested during the preoperational testing program to ensure its capability of performing its intended function. This change has

{ been submitted to the NRC by letter dated August 30, 1985 (ANPP-33309).

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! (29) Description of Change l A change was made to the PVNGS training program for the general employee retraining for non-licensed personnel. The change requires non-licensed plant personnel to receive retraining (general employee training) on an annual basis but not to exceed a period of 15 months. The previous

training program description required retraining on an annual basis.

1 Summary of Safety Evaluation This change to the general employee training program allows for a three month grace period for the annual retraining. This change does not I affect the ability of the training program to comply with the

{ recommendations of ANSI /ANS-3.1-1978. Thus, PVNGS Technical Specification 6.4.1 is met and the change does not involve an unreviewed .

safety question. This change is also in agreement with the PVNGS Safety

Evaluation Report _ (NUREG-0857) which states that general employee retraining shall be conducted at ' intervals not to exceed two years.

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i This change has been submitted to the NRC by letter dated August 30, 1985 (ANPP-33306).

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(30) Description of Change The FSAR was revised to reflect changes that were made to the ANPP organization as well as changes that were made to position titles.

Additionally, the FSAR was revised to require that members of the Independent Safety Engineering Group (ISEG) have at least two years average level of nuclear power plant experience.

Summary of Safety Evaluation This change does not involve an unreviewed safety question. The organizational changes that were made do not decrease plant safety.

These organizational changes were made to achieve a more efficient operating organization for PVNGS. The change that was made to the ISEG experience requirement is in accordance with PVNGS Technical Specification 6.2.3.2 which requires the ISEG personnel to have at least two years of professional level experience in his field. This change has been submitted to the NRC by letter dated August 30, 1985

( ANPP-33314) .

(31) Description of Change Revised the Quality Assurance program regarding the effect of nonconforming items on the performance of preoperational tests. The change requires the evaluation of nonconforming items prior to the initiation of the preoperational test. Nonconforming items which are determined to have an effect on the test perf ormance will be resolved prior to the affected step of the test.

Summary of Safety Evaluation The purpose of this change is to allow for more flexibility in the dispositioning of nonconformances that affect the preoperational test program. The QA program originally stated that nonconformances will be resolved prior to the start of the preoperational test. The revised program allows for evaluation of the nonconformance prior to the start of the preoperational test. If it is determined that the nonconformance will not have an impact on the outcome of the test, the test will be performed prior to final resolution of the nonconformance. If it is determined that the nonconformance will impact the outcome of the test, 1

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the nonconformance will be resolved prior to the affected step of the preoperational test. This change meets the requirements of 10 CFR 50, Appendix B and Regulatory Guide 1.28 because procedures are in place to evaluate the effect of nonconformances on preoperational testing.

Th ere fore , this change does not. constitute an unreviewed safety question. This change has been submitted to the NRC by letter dated September 27, 1985 (ANPP-33581).

(32) Description of Change Change to the power ascension testing program to remove the Reactor Power Cutback System (RPCS) from the test method and acceptance criteria for certain power ascension tests.

Summary of Safety Evaluation The RPCS is a non-safety related system that is not required in order to prevent an accident or to mitigate the consequences of an accident. The function of the RPCS is to produce a step reduction in reactor power following a large load rejection, turbine trip, or a loss of one of the main feedwater pumps. The accident analyses for these events do not credit the use of the RPCS to mitigate the event. As an example of this, the accident analysis for a turbine trip relies on the reactor tripping on high pressurizer pressure. There are other safety related systems, such as the plant protection system, that are relied upon for plant safety af ter these types of events. Therefore , this change does not constitute an unreviewel safety question. This change has been submitted to the NRC by letter dated August 23, 1985 (ANPP-33254).

(33) Description of Change This change involves a revision to procedures as described in the FSAR.

Specifically, the responsibility for validation of Nonconformance Reports (NCRs) has been transferred from the Project Quality Assurance Manager to the Project QC Engineer.

Summary of Safety Evaluation This change does not involve an unreviewed safety question. NCRs will continue to be validated in accordance with applicable procedures. This change only changes the responsibility for NCR validation. Thus, the PVNGS QA program continues to meet the requirements of 10 CFR 50, Appendir B. This change has been submitted to the NRC by letter dated August 30, 1985 (ANPP-33311).

(34) Description of Change This change is a minor editorial change to the FSAR to correct an inconsistency in a cross-reference table on FSAR Figure 3.6-4.

Summary of Safety Evaluation This FSAR change is an editorial change and makes no physical modifications to the facility. FSAR Figure 3.6-4 incorrectly refarenced Figure 3.6-3 for break number 2206. Break number 2206 is actually shown on Figure 3.6-2. The FSAR was revised to present the correct reference. This change has been submitted to the NRC by letter dated August 30, 1985 (ANPP-33312).

(35) Description of Change The existing FSAR Section 11.3.1.1 states that vaste gas will be held for 45 days prior to release. This has been changed to state that the gaseous radwaste system has been sized to provide the capability of holding radioactive gas for a 45 day decay period.

Summary of Safety Evaluation There is no requirement to hold gaseous radwaste for a period of time l prior to release. The only criterion that must be considered is that the tank must be sampled prior to release to aid in the determination of the release rate. The release rate is determined so that the limits of de PVNGS Technical Specifications and the applicable NRC regulations are not exceeded. Additionally, there is no increase in the probability or the consequences of an accident as the holdup period has no effect on the waste gas decay tank rupture accident scenario. This change does not involve an unreviewed safety question.

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l (36) Description of Change Section 16 of the PVNGS FSAR originally contained the preliminary Technical Specifications. Upon the issuance of the Operating License for PVNGS Unit 1, this section was deleted and all references to the original Section 16 were revised to reference the final Technical Specifications which were included as an attachment to the Operating License.

Summary of Safety Evaluation This change is administrative and does not involve an unreviewed safety 1 question. The change was made so that the FSAR would reference the appropriate facility Technical Specifications. This change has been j previously submitted to the NRC by letter dated August 30, 1985 (ANPP-33312) .

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1 (37) Description of Change The FSAR was changed to show that the turbine bypass valves are designed to fully close within 5 seconds. The previous FSAR description stated that the valves were designed to fully close within 1 second.

Summary of Safety Evaluation The turbine bypass system is a non-safety related system which is not credited for. accident mitigation or safe shutdown in the accident analyses. The atmospheric dump valves provide the safety related methed of relieving steam from the secondary system for decay heat removal.

Additionally, the closure times for the steam bypass valves are not specified in the PVNGS Technical Specifications. This change does not

, involve an unreviewed safety question and has been submitted to the NRC by letter dated August 30, 1985 (ANPP-33313).

(38) Description of Change The facility design was changed to allow for a nominal 3 inch gap between the containment liner plate and the internal structures for l

seismic displacement allowances. Gaps smaller than 3 inches may be I

permitted based upon the results of a case-by-case evaluation. The j previous design (as described in the FSAR) required a 6 inch seismic gap.

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f Summary of Safety Evaluation This change does not involve an unreviewed safety question. The change still maintains an adequate separation between the containment liner plate and internal structures under postulated safe shutdown earthquake conditions. The adequacy of the 3 inch gap was determined based upon a comparison of the maximum anticipated motions and the 3 inch gap. This change has been submitted to the NRC by letter dated August 30, 1985 (ANPP-33306).

(39) Description of Change Section 3.8.1.2.2 of the PVNGS FSAR was amended to reference the correct code for containment design. The existing FSAR section incorrectly referenced ACI 301-72 for containment design. This referenced code is applicable to the construction of the containment. The correct code for the containment design is ACI 318-71.

Summary of Safety Evaluation This change does not involve an unreviewed safety question as it is an i

editorial change. The correct standard (ACI 318-71) was used in the design of the containment and is properly referenced in the project Design Criteria Manual. This change has been submitted to the NRC by letter dated August 30,1985 (ANPP-33313).

(40) Description of Change The existing FSAR implied that the diesel generator fuel oil storage tanks are maintained at their 100% full level during normal operations.

The FSAR has been revised to clarify that the 100% full level isthe design capacity and not the normal tank inventory.

l Summary of Safety Evaluation Each diesel generator fuel oil storage tank is designed to provide inventory for the continuous operation of one diesel generator for 7 days plus 15% margin. 'Ihe PVNGS Technical Specifications (Te ch. Spec.

3.8.1.1) as well as other regulatory guidance only requires the inventory in the tanks to be maintained at the level for 7 days of diesel operation. This is approximately 71,500 gallons of fuel oil and

corresponds to an 80% indicated level in the tanks. Thus, the tanks are maintained at a minimum 80% level which is sufficient to ensure that the diesel generators are capable of performing their safety related function. This change does not constitute an unreviewad safety question. This change has been submitted to the NRC by letter dated August 30, 1985 (ANPP-33313).

(41) Description of Change The FSAR was revised to change the procedure for initial criticality.

Specifically, the change will allow PVNGS Units 2 and 3 to achieve i

initial criticality by either withdrawal of the last regulating group of CEAs or by RCS boron dilution. In either of these two methods, the last regulating group of CEAs will be used to control the chain reaction following criticality.

Summary of Safety Evaluation Initial criticality for PVNGS Units 2 and 3 will be achieved in accordance with approved plant procedures. This change simply permits initial criticality to be achieved by an alternate method from the current method of dilution which is described in CESSAR Section 14.2.10.2. This change will not introduce a configuration any different than those evaluated previously in the FSAR. Therefore, this change does not constitute an unreviewed safety question and this change has been previously submitted to the NRC by letter dated August 30, 1985

( ANP P-33315) .

(42) Description of Change Table 2.5-17 of the FSAR was changed to correct a minor typographical error. Specifically, the units should be 1bs/ft per foot of wall height , for horizontal backfill. The previous FSAR table presented the units as 1bs/f t per foot of wall height.

Summary of Safety Evaluation This change is to correct the units associated with the horizontal backfill pressure in FSAR Table 2.5-17. This change only involves the correction of a minor typographical error and does not involve an unreviewed safety question.

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The Tables in Section 5.2, 5.3, and. 5.'4 of the PVNGS FSAR are changed to revise the fracture toughness data for all three units and to correct discrepancies in the data. ,

Y Summary of Safety' Evaluation The specifications of the equipmen't and component's ' listed in Sections 5.2, 5.3, and 5.4 of the PVNGS FSAR have not changed. The equipment and components will function as previo'usly analyzed. Therefore, this change does not constitute an unreviewed' safety question. This change was made in response to an NRC request and was submitted to the NRC by letter dated August 30, 1985 (ANPP-33305).

(44) Description of Chan:le A change was made to the organi::stional responsibilities for the startup test program. Specifically, the position of Shif t Test ; Director (STD) has been deleted. The STD respcssibilities are now assigned to the Responsible Engineers who are djrectly involved with the testing.

Susmary of Safety Evalisation This change is an 'organiza tional change and alters some of the responsibilities that are described in Section 14.2 of the PVNGS FSAR.

The changes do not constitute an unreviewed safety question. The changes consolidate the responsibility for test conduct with those personnel who nre directly responsible for ' the performance of the tests. Additionally, this change has been transmitted to the NRC by letter dated October 22, 1985 (ANPP-33766).

(45) Description of M a Section 9.5.4.4 (: th ; VNGS FSAR was revised. to be consistent with the PVNGS Technical Specifications regarding the quality of diesel generator fuel oil. The specifications for the API gravity, specific gravity, and the absolute specific gravity for the fuel oil were changed to be consistent with the( requirements of Technical Specification 3.8.1.1.

Additionally, the numbers for the high and low heating values for the fuel oil are deleted from the FSAR since this specification is not required by the applicable regulatory criteria.

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1 Summary of Safety Evaluation The emergency diesel generator fuel oil specifications are designed to j ensure that the diesel generator will not malfunction in an accident situation as a result of poor quality fuel oil. PVNGS currently tests l

and procures the fuel ~ oil in accordance with the PVNGS Technical Specifications, ASTM D-975, and Regulatory Guide 1.137. his change to 4

the PVNGS FSAR has no effect on plant safety as the specifications of the fuel oil will continue to meet the guidelines of Regulatory Guide 1.137 for the fuel oil quality and testing. Additionally, the requirements of PVNGS Technical Specification 3.8.1.1 will be satisfied. Thus, there will be no decrease to the margin of safety as defined in the bases for that Technical Specification. This change to make the FSAR consistent with the Technical Specifications does not involve an unreviewed safety question.

(46) Description of Change the FSAR was revised to clearly state the qualification requirements for certain Quality Assurance (QA) personnel. The following personnel qualifications shall apply:

(i) personnel performing audits shall be qualified to ANSI N45.2.23; (ii) personnel performing material inspection shall be qualified to l ANSI N45.2.6; and (iii) personnel performing monitoring activities shall be qualified to l ANSI N45.2.23 or ANSI N45.2.6.

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Summary of Safety Evaluation This change does not decrease the effectiveness of the QA program and does not involve an unreviewed safety question. All audits or monitoring activities conducted at vendor's facilities will be conducted i

by trained personnel. This change clarifies the personnel qualifications that are applicable to each of the separate QA functions.

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(47) Description of Change Table 3.2-1 of the PVNGS FSAR was revised to describe the actual Quality Assurance (QA) program for the new fuel racks. This QA program does not fully meet the criteria of 10 CFR 50 Appendix B, but the program is sufficient to assure that the safety related function of the racks im satisfied.

Summary of Safety Evaluation The QA program under which the new fuel storage racks were fabricated does not fully meet the criteria of 10 CFR 50, Appendix B. However, since the only safety related function of these racks is to maintain the minimum surface to surface separation between assemblies, the QA program that was applied to the racks is acceptable. The design, specific components, and fabrication processes critical to the structural integrity of the new fuel storage racks are sufficient to ensure that the equipment will withstand the effects of the SSE and remain functional in accordance with Regulatory Guide 1.29.

(48) Description of Change Section 9B.2.11 cf the FSAR was revised to correct a minor typographical error. This section of the FSAR describes the safe shutdown equipment for specific fire areas. The specific change involves correcting the instrument tag number for the steam generator No. 1 wide range level instrument from J-SGA-LT-1113B to J-SGB-LT-1113B. This level instrument is a Train-B component.

Summary of Safety Evaluation This change is to correct a minor typographical error in the FSAR and does not involve an unreviewed safety question. This change does not impact the evaluations which were conducted for safe shutdown and does not impact plant safety.

(4 9) Description of Change The FSAR was revised to eliminate the references to chlorine detectors at PVNGS. The chlorine detectors were previously removed from the FSAR, but all of the FSAR references to the detectors were not removed.

Summary of Safety Evaluation The purpose of this change was to eliminate all references to the chlorine detectors at PVNGS. These detectors were eliminated from the PVNGS design because there is no onsite storage of liquid or gaseous chlorine at PVNGS. This eliminates the potential for chlorine contamination in the atmosphere which could be drawn into the control room by the HVAC supply system. The NRC staff had previously approved of the removal of the chlorine detectors from the PVNGS design. This approval is documented in Supplement No. 3 to the PVNGS Safety Evaluation Report.

(50) Description of Change FSAR Figures 9B-36 and 9B-37 were revised to correctly show the extent of wet pipe sprinkler coverage in the turbine buildings at elevations 100 feet and 140 feet.

Summary of Safety Evaluation This change did not result in a physical change to the facility. The FSAR was revised to reflect the as-built condition of the facility. 'Ihe safety of the plant is not compromised as a result of this change as no equipment which is relied upon for acciuent mitigation or for the safe shutdown of the plant is located in the turbine building. Therefore, this change does not involve an unreviewed safety question.

(51) Description of Change The initial test program for PVNGS Units 2 and 3 was revised. The following is a description of each of the changes: (1) For Units 2 and 3, the CEDM performance testing which is described in CESSAR Section 14.2.12.3.4 will be conducted at hot, zero power conditions only. (2)

For Units 2 and 3, only the regulating CEA groups will be measured for CEA group worth. If this measurement of group worth does not agree with predicted values, the test program will be expanded to include the measurement of shutdown group uorths. This testing is described in Se ction 14.2.12.4.4 of CESSAR. (3) The test methcd for the Variable T test, which is described in CESSAR Section 14.2.12.5.2, was AVG changed for Units 2 and 3 to allow for flexibility in the test method.

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The isothermal temperature coefficient can be measured by CEA movement

or by adjusting the RCS boron concentration. The power coefficient can be measured by CEA adjustment or by balancing core average temperature. -

(4) For Units 2 and 3, the testing of all moveable incore detectors will not be conducted. (5) For Units 2 and 3, the control systems checkout test will be conducted at 50% and 100% power. This is a change to the previous commitment to perform the tests at 50% and 80% power. (6) The turbine trip test will not be conducted for Units 2 and 3.

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Summary of Safety Evaluation i The primary purpose for these changes to the initial test program is due a

j to the fact that Units 2 and 3 are follow-on units. Thus, all of the testing that - was conducted on the first of-a-kind System 80 unit does not have to be repeated for the follow-on units. All three PVNGS units are designed to be essentially the same so that the performance of the plants will be very similar. These changes do not involve an unreviewed safety question as the plants will be tested sufficiently to ensure that the plant is capable of responding to off-normal conditions. These changes have been previously submitted to the NRC by letter da'-d August 30,1985 (ANPP-33315).

(52) Description of Change This change to the facility involves changing the design pressure of the ,

Equipment Drain Tank (EDT) from 60 psig to 30 psig.

Summary of Safety Evaluation

, The original design pressure of the EDT was specified in CESSAR Section 9.3.4 as 60 psig. The PVNGS design does not meet this design pressure due to the fact that the downstream side of the pressure regulating ,

valve (CH-831) cannot meet a pressure requirement greater than 30 psig.

This change does not involve an unreviewed safety question as the function of the EDT is to serve as a collection point for reactor coolant quality water which may be recovered from drains or leakage from systems outside of the containment building. The failure of this tank

would result in the inability to recover this reactor coolant quality 4

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water and the water would be collected in the auxiliary building drains system and transferred to the liquid radwaste system. The EDT does not serve any accident mitigation function.

(53) Description of Change The remote seismic monitors were removed from the field. Section 2.5 of the PVNGS FSAR was revised to reflect the removal of the remote seismic monitors.

Summary of Safety Evaluation The remote seismic monitors were installed in 1975 in order to collect seismic information on the PVNGS region. This information is required by 10 CFR 100 to characterize the seismic activity of the region. There is a permanent plant seismic monitoring system, which is required by the Technical Specifications, to gather information on seismic activity.

The remote seismic monitoring, which was removed, did not serve any accident monitoring or mitigation function and is not required to safely shutdown the unit. The remote seismic monitoring system had fulfilled its information gathering function of collectir.g data on seismic activity in the region. Therefore, the removal of the seismic monitoring instru;aentation does not involve an unreviewed safety questien.

(54) Description of Change The previous FSAR describes an in-line type Post-Accident Sampling System (PASS) which is fully automated to provide analysis results of post-a ccident samples from the RCS hot legs, containment sumps, auxiliary building sumps, PCCS A & B pumps mini-flow lines for liquid samples and from the containment hydrogen control system for gas samples. The facility and the FSAR have been modified to describe a grab sample type PASS which provides samples for post-accident analysis in the hot lab. The sources for the post-accident liquid samples are the RCS hot leg, letdown line, containment radwaste sumps, auxiliary building sumps, and the ECCS A & B pumps mini-flow lines. The source for the post-accident grab sample of the containment atmosphere is the containment hydrogen control system.

Summary of Safety Evaluation This change does not constitute an unreviewed safety question. The PASS is not a safety system used in mitigating the consequences of an accident. The purpose of the PASS is to provide samples for post-accident analysis in order to quantify radionuclides that are indicators of core damage, hydrogen levels in containme.nt atmosphere, concentrations of dissolved gas, chlorides, and boron for liquid samples. The PASS is not credited in any of the accidents analyzed in the FSAR nor does it provide a function for maintaining the plant 'na i

j safe operating condition. Additionally, this change has been submitted to the NRC by letter dsted September 26, 1985 (ANPP-33573).

(55) Description of Change This change involves a change to the project procedures which will permit the use of the operations phase Quality Assurance (QA) program for procurement during the remainder of the construction phase.

Summary of Safety Evaluation This change does not involve an unreviewed safety question as the change still requires that all procurement activities be performed in conformance with QA procedures and controls. Additionally, this change will not decrease the effectiveness of the QA program since the revised program will be implementing portions of a previously approved QA program for operations phase activities. This change has been submitted to the NRC by [[letter::05000528/LER-1985-048, Corrected LER 85-048-00:on 850819,ECCS Subsystem RQ & Bp Declared Inoperable Due to Noncompliance W/Surveillance Requirement 4.5.2.g.1.Temporary Change Initiated to Permit Testing of ECCS Throttle Valves|letter dated September 18, 1985]] (ANPP-33515).

(56) Description of Change The PVNGS containment analyses were revised and . Incorporated into the FSAR. The revised analyses were necessary to reflect reduced containment spray pump flows, as-built piping sizes, and PVNCS plant specific changes as discussed in ANPP-32401 dated April 15, 1985.

Summary of Safety Evaluation The plant specific changes were evaluated by the re performance of the j containment analyses. The results of the revised analyses are within I the acceptance criteria. The new analyses verify that the existing

margin of safety as defined in the PVNGS Technical Specifications bases is maintained. Therefore, this change does not involve an unreviewed safety question. Additionally, the revised analyses have been submitted for NRC review by letter dated September 30,1985 (ANPP-33610).

(57) Description of Change The PVNGS Main Steam Line Break (MSLB) analyses were revised and incorporated into the FSAR. The revised analyses were necessary to demons trate the acceptability of various plant changes which were discussed in a previous letter to the NRC (refer to ANPP-32401 dated April 15, 1985).

Summary of Safety Evaluation The plant specific changes were evaluated by the re performance of the MSLB analyses. The results of the revised analyses are within the applicable NRC acceptance criteria and the analyses verify that the margin of safety, as defined in the bases of the PVNGS Technical S pecifications, is maintained. Therefore, this change does not involve an unreviewed safety question. Additionally, the revised analyses have been submitted for NRC review by letter dated September 30, 1985 (ANPP-33611) .

(58) Description of Change The fire protection section of the FSAR was revised to reflect the as-built condition of the plant. The specific changes are as follows:

(1) The CO 2 fire extinguisher shown on the east wall of fire area SA (FSAR Figure 9B-9) is actually located on the north end of the room adjacent to the HVAC chase; (2) The hose length at hose station #44 has been increased from 125 feet to 15 0 feet; (3) An exception was identified in FSAR Table 98.3-1 to identify the fact that the isolation valves to the preaction sprinkler heads at two missile doors in the corridor building are not electrically supervised; (4) The fire door on the east side of fire zone 30B is not shown on FSAR Figure 9B-18; (5)

Existing fire extinguishers are not shown in fire zones 37C and 37D; (6)

The deluge valve shown on FSAR Figure 9B-22 in fire zone 52B is a maintenance isolation valve and not a deluge valve; (7) The symbol for a hose station was added to FSAR Figure 9B-11 for hose station #93.

4 Summary of Safety Evaluation The purpose of these changes was to update the FSAR to reflect the as-built condition of the plant. These changes do not result in an unreviewed safety question as none of the changes result in the probability of occurrences or the consequences of an accident being increased from that previously evaluated in the FSAR.

(59) Description of Change

. The fire protection description contained in the FSAR was revised to

reflect the as-built condition of the plant. The specific changes are as follows

(1) All preaction systems, with the exception of the turbine driven auxiliary feedwater pump room, are actuated by single 4

zone products-of-combustion fire detectors; (2) The automatic preaction

] sprinkler system for those systems protecting electrical cable trays is initiated by either line type heat detectors or by products-oi'-

j combustion detectors.

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] This change does not involve an unreviewed safety question. The probability of occurrence of an accident remains unchanged since the system remains as a preaction system. Specifically, inadvertent actuation is not probable since- both the detector and the sprinkler heads must actuate in order for flow to be delivered to the affected

< area.

(60) Description of Change The FSAR was revised to clarify the radiation protection activities that are subject to the requirements of the operational Quality Assurance (QA) program. This clarification supplements the information contained in the body of FSAR Table 3.2-1 pertaining to radiation equipment and activities.

Summary of Safety Evaluation The change states that activities and equipment related to the calibration of radiation protection and chemistry equipment are subject to the applicable requirements of the operations phase QA program. This change does not involve an unreviewed safety question as the change does not reduce any previous commitments to apply QA program requirements to radiation monitoring equipment, the radiation monitoring system, or the radiation exposure management system. Additionally, the QA requirements that are applied to radiation protection equipment are sufficient to ensure that the equipment will be available to perform its intended safety function.

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