DCL-16-015, Response to NRC Request for Additional Information Regarding License Amendment Request 15-03, Application of Alternative Source Term.

Response to NRC Request for Additional Information Regarding License Amendment Request 15-03, Application of Alternative Source Term.

ML16032A603
Person / Time
Site:	Diablo Canyon
Issue date:	02/01/2016
From:	Welsch J Pacific Gas & Electric Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
DCL-16-015
Download: ML16032A603 (54)

Text

Pacific Gas ,and Ele,ctrtc Company James M. Welsch Diablo Canyon Power Plant Vice President, Nuclear Generation P.O. Box 56 Avila Beach, CA 93424 805.545.3242 E*Mail: JMWl @pge.com February 1, 2016 PG&E Letter DCL-16-015 U.S. Nuclear Regulatory Commission 10 CFR 50.90 ATTN: Document Control Desk Washington, DC 20555-0001 Docket No. 50-275, OL-DPR-80 Docket No. 50-323, OL-DPR-82 Diablo Canyon Units 1 and 2 Response to NRC Request for Additional Information Regarding License Amendment Request 15-03, "Application of Alternative Source Term"

References:

1. PG&E Letter DCL-15-069, License Amendment Request 15-03, "Application of Alternative Source Term," dated June 17, 2015 (ADAMS Accession No. ML15176A539)

2. PG&E Letter DCL-15-105, Supplement to License Amendment Request 15-03, "Application of Alternative Source Term," dated August 31, 2015 (ADAMS Accession No. ML15243A363)

3. E-mail from NRC Project Manager Siva P. Lingam, "Diablo Canyon 1 and 2 - Requests for Additional Information for License Amendment Request 15-03 to Adopt the Alternative Source Term per 10 CFR 50.67 (TAC Nos. MF6399 and MF6400),"dated December 2, 2015

Dear Commissioners and Staff:

License Amendment Request (LAR) 15-03, "Application of Alternative Source Term" was submitted by Pacific Gas and Electric (PG&E) Letter DCL-15-069 (Reference 1) and supplemented by PG&E Letter DCL-15-1 05 (Reference 2).

In Reference 3, the NRC Radiation Protection and Consequence Branch (ARCB) requested additional information required to complete the review of LAR 15-03.

PG&E's responses to the ARCB Staff's questions are provided in the Enclosure.

This information does not affect the results of the technical evaluation or the no significant hazards consideration determination previously transmitted in References 1 and 2.

A member of the STARS (Strategic Teaming and Resource Sharing) Alliance Callaway

• Diablo Canyon

• Palo Verde*

• Wolf Creek

I~

Document Control Desk PG&E Letter DCL-16-015 February 1, 2016

• Page 2 PG&E makes no regulatory commitments (as defined by NEt 99-04) in this letter.

This letter includes no revisions to existing regulatory commitments.

If you have any questions, or require additional information, please contact Mr. Hossein Hamzehee at (805) 545-4720.

I state under penalty of perjury that the foregoing is true and correct.

Executed on February 1, 2016.

Sincerely,

~~~

James M. Welsch Vice President, Nuclear Generation e 1d7/4418/50705089 Enclosure cc: Diablo Distribution cc/enc: Marc L. Dapas, NRC Region IV Administrator Siva P. Lingam, NRR Project Manager Gonzalo L. Perez, Branch Chief, California Department of Public Health Binesh K. Tharakan, Acting NRC Senior Resident Inspector A member of the STARS (Strategic Teaming and Resource Sharing) Alliance Callaway

• Diablo Canyon

• Palo Verde*

• Wolf Creek

Enclosure PG&E Letter DCL-16-015 PG&E Response to NRC Request for Additional Information (RAI) Regarding License Amendment Request 15-03, "Application of Alternative Source Term" NRC ARCB-RAI-1 In the Enclosure to the application dated June 17, 2015 it states:

Full implementation of [alternate source term] AST for DCPP Units 1 and 2 does not include revising ... NUREG-0737 responses associated with shielding and vital area access.

However, RG 1.183, Regulatory Position 4.3, "Other Dose Consequences," states that:

The guidance provided in Regulatory Positions 4.1 and 4.2 should be used, as applicable, in re-assessing the radiological analyses identified in Regulatory Position 1.3.1, such as those in NUREG-0737. Design envelope source terms provided in NUREG-0737 should be updated for consistency with the A ST. In general, radiation exposures to plant personnel identified in Regulatory Position 1.3.1 should be expressed in terms of [total effective dose equivalent]

effective TEDE.

In evaluating the submittal, the NRC staff could not determine how RG 1. 183, Regulatory Position 4. 3 has been assessed for DCPP. Please provide additional information describing how Regulatory Position 4. 3 has been assessed for DCPP.

PG&E Response In accordance with Regulatory Guide (RG) 1.183, Regulatory Position 4.3 and summarized below, Diablo Canyon Power Plant (DCPP) has assessed the effect of reassessing the radiological analyses performed in support of NUREG-0737 (and identified in Regulatory Position 1.3.1 as also affected by the accident source term) using the dose calculation methodology identified in Regulatory Positions 4.1 and 4.2 of RG 1.183.

Post-Accident Access Shielding (NUREG-0737. 11.8.2)

DCPP shielding study, which was developed in response to NUREG-0737, Item 11.8.2, was submitted to the NRC via PG&E Letter DCL-84-260, "Radiation Shielding Review,"

Revision 3, dated July 12, 1984. This report documented the estimated radiation exposure to plant personnel performing vital missions in support of accident mitigation and safe shutdown following a Loss-of-Coolant Accident (LOCA).

The source terms used in the referenced plant shielding study were based on traditional TID-14844 assumptions; specifically those outlined in NUREG-0737, Item 11.8.2. The nuclide release fractions specified in the Alternate Source Term (AST) methodology

Enclosure PG&E Letter DCL-16-015 outlined in RG 1.183 Revision 0, differ from those outlined in TID-14844/ NUREG-0737.

The difference in the release fractions has the potential to affect the dose rates in vital areas where piping containing post-LOCA sump fluid are located.

As discussed in NUREG-0933, Generic Issue 187, a study conducted by Sandia National Labs, "Evaluation of Radiological Consequences of Design Basis Accident at Operating Reactors Using the Revised Source Term," dated September 28, 1998, showed that exposure to containment atmosphere sources developed based on traditional source term methodology and AST methodology produced similar integrated doses. This report also showed that the integrated AST doses from exposure to post-LOCA sump fluid did not exceed those based on TID-14884 assumptions until42 days after an event at a pressurized water reactor (PWR).

Based on the above study it can be concluded that the differences in the release fractions associated with AST methodology would have little impact on the local dose rates during the 30-day post-LOCA mission time. Since the local dose rates are not expected to be significantly impacted by AST during the first 30 days following a LOCA, the conclusions of the shielding study with respect to operator exposure would not significantly change by expressing the mission doses in terms of total effective dose equivalent (TEDE).

The doses to the operator in the Control Room (CR) and Technical Support Center (TSC) were recalculated as part of the AST application and are presented in AST LAR 15-03 (Reference 1) in terms of TEDE.

Post-Accident Sampling Capability (NUREG-0737. 11.8.3)

Technical Specification (TS) Amendments 149 (Unit 1) and 149 (Unit 2) eliminated the requirements to have and maintain the Post-Accident Sampling System.

The post-accident sampling ability has been retained as a potential resource for data, to be used at the discretion of Chemistry personnel. As discussed above, should this ability be exercised, the previously estimated local dose rates near the sample panel are not expected to be significantly impacted by AST during the first 30 days following a LOCA, and the operator dose due to exposure would not significantly change by expressing the mission doses in terms of TEDE.

Accident Monitoring Instrumentation (NUREG-0737, II.F.1)

Post-accident monitoring instrumentation is available to measure various plant parameters. This instrumentation is designed to detect and remain operable under postulated design basis conditions. Although the more realistic AST source term would potentially involve a larger dose to equipment exposed to sump water over long periods of time, the conclusions of NUREG-0933, Generic Issue 187, which specifically addresses this issue, stated that there would be no discernable risk reduction associated with modifying the design basis for equipment qualification to adopt AST.

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Enclosure PG&E Letter OCL-16-015 Therefore, the ability of the accident monitoring instrumentation to meet NUREG-0737 requirements is not impacted by the AST.

Leakage Control (NUREG-0737, 111.0.1.1)

Section 5.5.2 of the TS establishes a program to reduce leakage from primary coolant sources outside containment. As discussed in Section 2.3, Item 5, Attachment 4 of LAR 15-03, the engineered safety feature (ESF) system leak testing procedures that are invoked by TS 5.5.2 will be updated to reflect administrative acceptance criteria that ensure that the allowable leakage is at least a factor of two less than that assumed in the dose consequences analyses performed in support of AST implementation.

Emergency Response Facilities (NUREG-0737, III.A.1.2)

Emergency response facilities are shared by both Units 1 and 2. The TSC and the Operational Support Center (OSC) are co-located. The dose consequences in the TSC/OSC have been evaluated using AST methodology. The Emergency Operations Facility (EOF) is located outside of the 10 mile Emergency Planning Zone (EPZ).

Control Room Habitability (NUREG-0737, 111.0.3.4)

Radiological analyses were performed to meet the requirements of Item 111.0.3.4 of NUREG-0737. These analyses ensure that the CR operators are adequately protected against the effects of accidental release of radioactive gases and that the plant can be safely operated or shutdown under design basis conditions. As part of the AST implementation, the radiological analyses supporting CR habitability following design basis accidents were recalculated in terms of TEOE dose conversion factors and acceptance criteria.

NRC ARCB-RAI-2 In Enclosure Attachment 4, Section 7.2.6 it states:

To address the existing licensing basis, a TEDE dose is estimated for operator access to the control room. Because RG 1. 183 does not provide guidance on determining the egress and ingress to the control room following an accident, the same inputs used to estimate the current licensing basis values for access to the control room, along with the associated dose estimate presented in the UFSAR, are used to determine the TEDE dose estimate for ingress/egress.

In addition, Enclosure Table 1 and Enclosure Attachment 4 Table 8.1-1 state that the control room total effective dose equivalent presented for the loss of coolant accident represents the operator dose due to occupancy which is 3. 7 rem and that the value shown in parenthesis represents that portion of the total dose reported that is the contribution of direct shine from contained sources/external cloud which is 0. 7 rem.

Furthermore, they stated that the dose to the control room operator during routine 3

Enclosure PG&E Letter DCL-16-015 access for the 30 day duration of the accident is discussed in Enclosure Attachment 4, Section 7.2.6 and summarized in the text of Enclosure Attachment 4, section 8.0 which is 0. 037 rem.

10 CFR 50. 67, "Accident source term," (b)(2) states:

The NRC may issue the amendment only if the applicant's analysis demonstrates with reasonable assurance that:

(i) .. .

(ii) .. .

(iii) Adequate radiation protection is provided to permit access to and occupancy of the control room under accident conditions without personnel receiving radiation exposures in excess of 0. 05 Sv (5 rem) total effective dose equivalent (TEDE) for the duration of the accident.

In order to meet 10 CFR 50. 67, the radiation dose for accessing the control room must be evaluated. PG&E proposes to use the same inputs used to estimate the current licensing basis values for accessing the control room. However, the AST uses a different source term than that in the current licensing basis. Therefore, please provide an analysis of the radiation dose received from accessing the control room that reflects the new source term proposed with implementing the A ST.

PG&E Response As noted in Section 7.2.6 of Attachment 4 of LAR 15-03, the radiation dose received by the operator during the outbound and inbound excursions from the CR to the site boundary during the 30-day period post-LOCA is developed taking into consideration:

a) Use of the new source term proposed with implementation of AST b) The additional post-LOCA fission product release pathways assessed in LAR 15-03.

The assessment presented in Section 7.2.6 of Attachment 4 of LAR 15-03 takes into consideration the following:

a) Transit to and from the CR is expected after the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> following the accident by which time the airborne levels inside containment has reduced

• significantly due to the use of active fission product removal mechanisms such as containment sprays, and by radioactive decay, b) The operator is protected from radioactive ESF fluids by the shielding provided by the buildings that house such equipment, and c) Routine ingress/egress to the CR during the 30-day period following a LOCA falls into the mission dose category as discussed in NUREG-0737, II. B. 2.

NUREG-0737, Item II. B. 2, states that leakage of systems outside containment need not be considered as potential sources.

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Enclosure PG&E Letter DCL-16-015 In accordance with DCPP original licensing basis, radiation exposures to personnel during egress and ingress (i.e., during routine access to the CR for the duration of the accident) could result from the following sources:

a) Airborne activity in the containment leakage plume b) Direct gamma radiation from fission products in the containment structure As noted in Section 7 .2 .6 of Attachment 4 of LAR 15-03, no additional sources need to be addressed to assess the operator dose due to ingress/egress to the CR since all of the additional post-LOCA fission product release pathways assessed with this application fall either into the category of (a) are terminated prior to 24hr after accident initiation (i.e., containment vacuum relief and RHR pump seal failure), or (b) "releases due to leakage of systems outside containment," and therefore need not be addressed per the guidance provided in NUREG-0737, II. B. 2 (i.e., refueling water storage tank (RWST) and miscellaneous equipment drain tank (MEDT) releases).

Since the configuration and shielding associated with the containment structure and CR has not changed with this application, the only change is in the source term.

The effect of the change in source term on the original licensing basis dose to CR personnel during egress/ingress reported in the DCPP Updated Final Safety Analysis Report (UFSAR), Table 15.5-33, and the subsequent development of the ultimate TEDE dose, is addressed using scaling techniques, and discussed in detail in Section 7.2.6 of Attachment 4 of LAR 15-03.

NRC ARCB-RAI-3 Please provide the RAD TRAD input and output files, in electronic format, for each of the AST DBAs described in the LAR.

PG&E Response In response to PG&E request, and as agreed to in the e-mail from Siva Lingam (NRC) to K. Schrader (PG&E) dated December 29, 2015, the RADTRAD input and output files were provided for NRC review during the January 12-14, 2016, NRC Audit of the calculations supporting DCPP AST LAR 15-03. This approach is in lieu of submitting the RADTRAD input and output files (considered to be proprietary by the vendor) with a request for withholding from the public via this response.

NRC ARCB-RAI-4 Please confirm that the failed fuel percentage (1 0%) stated in the Enclosure Attachment 4 Table 4.3-1 note, which is applied to the control rod ejection accident and the locked rotor accident, is equivalent to 10% of the rods in the core as opposed to 10% of the rods in an assembly.

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Enclosure PG&E Letter DCL-16-015 PG&E Response PG&E confirms that the failed fuel percentage ( 10 percent) stated in the Enclosure Table 4.3-1 note, which is applied to the control rod ejection accident and the locked rotor accident, is equivalent to 10 percent of the rods in the core.

NRC ARCB-RAI-5 DELETED.

NRC ARCB-RAI-6 Enclosure Attachment 4, Section 6. 2, ((Direct Shine Dose from External and Contained Sources," states:

CB&I S&W Inc. [Chicago Bridge and Iron Stone and Webster Incorporated] [A CB&I Company] point kernel shielding computer program SW-QADCGGP is used to calculate the deep dose equivalent (DOE) in the control room, [technical support center] TSC and at the [exclusion area boundary] EAB due to external and contained sources. The calculated DOE is added to the inhalation (CEDE) and the submersion factors are used and the geometry models are prepared to ensure that un-accounted streaming/scattering paths were eliminated. The dose albedo method with conservative albedo values is used to estimate the scatter dose in situations where the scattering contributions are potentially significant.

[American National Standards Institute/American Nuclear Society] ANSIIANS 6.1.1-1977 ((neutron and gamma-ray flux-to-dose-rate factors" (Reference 31) is used to convert the gamma flux to the dose equivalent rate.

Enclosure Attachment 4, Section 7.2. 7, ((Technical Support Center Dose," provides further detail on the direct shine dose to the technical support center from external and contained sources. Enclosure Attachment 4, Section 7. 2. 7, states:

CB&I S&W Inc. computer code PERC2 is used to calculate the dose to TSC personnel due to airborne radioactivity releases following a [loss of coolant accident] LOCA. The direct shine dose to an operator in the TSC due to contained or external sources resulting from a postulated LOCA is calculated using CB&I S&W Inc. point kernel shielding computer program SW-QADCGGP.

The post-LOCA gamma energy release rated [mega electron volt per second]

(Me Vlsec) and integrated gamma energy release [mega electron volt hour per second] (Me V-hrlsec) in the various external sources are developed with computer program PERC2.

In evaluating the LAR, the NRC staff could not thoroughly review and perform confirmatory calculations of the direct shine dose due to external and contained sources at the control room, TSC, low population zone (LPZ), and at the EAB from the 6

Enclosure PG&E Letter DCL-16-015 discussions described in the LAR. Therefore, please provide additional information in enough detail that will enable the NRC staff to be able to perform an independent calculation of the direct shine dose to the control room and the TSC from external and contained sources.

PG&E Response All Figures referenced in this response are included in enclosed Attachment 1. Figure 1 provides a plant layout and the general location of the CR and the TSC.

Figures 2 through 12 provide the layout arrangement of the CR, the location of radiation sources with respect to the CR, and pictorial representations of some of the shielding models used to simulate the plant configuration.

Figures 13 through 16 provide similar information for the TSC.

Control Room The CR is located in Plant Area "H" as shown on Figure 1. The overall plan view (Elevation (EI) 140ft) of the Unit 1 and Unit 2 shared CR is illustrated in Figure 2. The numbers in the rectangular boxes are plant room ID numbers and the numbers in the diamond boxes are plant door ID numbers. The overall dimension of the main CR is 118 ft by 70ft, and the room height is 20 ft. The Mechanical Equipment Room (MER) that serves the main CR is located at El154 ft-6 inch (in.) along column line L. The MER is included in the CR emergency ventilation system envelope. See Figure 6.

The post-LOCA sources addressed herein for direct shine include:

a) Airborne activity in containment b) Buildup of activity on the control room ventilation system (CRVS) filters due to accumulation of activity resulting from 5 airborne release pathways, i.e.,

containment leakage, ESF system leakage via the plant vent, ESF system leakage via the containment penetration areas, the residual heat removal (RHR) pump seal failure and MEDT releases. RWST releases are minimal compared to the MEDT releases; therefore, the activity accumulation in the CRVS filter due to RWST release is not addressed.

c) The activity concentration in the contaminated cloud outside the CR pressure boundary due to environmental releases via the 5 release pathways discussed above, i.e., containment leakage, ESF system leakage via the plant vent, ESF system leakage via the containment penetration areas, the RHR pump seal failure, and MEDT releases. Due to the reason outlined above for the CRVS filter, the cloud associated with the RWST release is not addressed.

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Enclosure PG&E Letter DCL-16-015 d) Recirculating sump water activity in piping and equipment.

e) Sump water activity in the RWST due to RWST back leakage.

Containment Shine:

The calculated dose from direct shine of activity airborne in the containment is about 0.6 percent of the total direct shine DOE dose. This dose contribution is small because (a) the CR outer wall facing the containment is made of 3 ft thick concrete, the roof of the CR is made of 3ft 4 in. thick concrete and there is significant slant path through the remaining 2ft thick concrete CR walls and floor (due to the geometry). The cylindrical wall of the containment is 3 ft-8 in. thick concrete; (b) the azimuth of personnel hatch that is located at the same elevation as the CR is such that there is negligible contribution to the direct shine dose in the CR; and (c) other large containment penetrations either don't face the CR or, as in the instance of a Main Steam (MS) line penetration, are below the CR elevation. Both the personnel hatch and MS line penetration are modeled using QAD-CGGP. For an understanding of the configuration, see Figure 3 (elevation view), Figure 8 (plan view of containment and CR, model has an axis shift of 180 degrees for modelling convenience), Figure 9 (QAD bulk shielding containment model), and Figure 10 (shows the QAD model of the 2ft thick concrete shield wall in front of the 115 in~ diameter personnel hatch).

CR Filter Shine:

The calculated dose from direct and scattered radiation due to activity buildup on the CR emergency filters is small, i.e., about 0.6 percent of the total direct shine dose.

This is due to several reasons (a) the filter is located in the MER about 23ft beyond column L, (b) the wall separating the MER and CR is 2ft concrete, (c) the penetrations in this wall along column L are 17 ft-3 in. above the CR floor, and (d) the largest of the six penetrations in the wall is 30 in. by 26 in. See Figures 6 and 7.

Cloud Shine:

Contaminated cloud shine through doors in the outer walls of the CR is the most significant direct shine dose contributor to personnel located in the CR; especially through doors 509 and 508 at column line H (see Figure 2). Airborne radioactivity concentrations in the stairwell between doors 508 and 509 are assumed to be the same as the concentrations in the environment outside resulting from all identified leakages (i.e., containment, ESF, MEDT and the passive component failure). The radioactivity concentrations in the external cloud are based on atmospheric dispersion factors applicable to a receptor located at the center of the CR boundary at roof level. Credit is taken for solid angle reduction through the door openings. Since doors 508 and 509 are adjacent to the Turbine Building (TB), no credit is taken for shielding offered by the TB (mostly corrugated sheet metal). The doorway along column L leads into the Auxiliary Building (AB) and concrete shielding in the AB is credited.

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Enclosure PG&E Letter DCL-16-015 Currently there is a large opening in the wall along column H that leads into Room 507 (briefing room), however, as noted in Section 2.2 of Attachment 4 and Commitment 1 in of LAR 15-03, prior toAST implementation, PG&E will install shielding material at this location equivalent to that provided by the CR outer walls. Currently, administrative controls are in place to restrict post-LOCA access to Room 507.

There are a significant number of small floor penetrations in the CR ranging from 4 in. by 4 in. to 12 in. by 16 in. or 8 in. by 20 in. The penetrations are mostly small and spread out so they do not combine to make an "effectively" large penetration. CR panels are located on top of most (if not all) of these cable penetrations, so standing on top of a penetration is unlikely. Two of the largest penetrations are modeled, i.e.,

a 12 in. by 16 in. and an 8 in. by 20 in. penetration. The CR operator direct and scattered dose from airborne activity in the cable spreading room below the CR is conservatively based on the bounding penetration centerline dose. About 15 percent of the total direct shine dose comes from these floor penetrations.

ESF Component Shine:

The direct shine dose from pipes and components carrying recirculating containment sump water is negligible. Listed below is the relevant information associated with the location and thickness of concrete shielding available between the CR and the piping I components carrying ESF fluids that are located in the vicinity of the CR.

As shown in Figure 3 and discussed below, the CR which is located in the AB has significant shielding between it and the closest ESF components that are carrying post-accident radioactive fluids.

Control Room - El140ft Cable Spreading Room - El128 ft Battery Room - El 115ft Switchgear Room - El1 00 ft ChernE Office I Lab- El 85ft CCW I Charging pumps - El 73ft Laundry- El 60ft The pipe penetrations in the containment wall in the containment penetration area adjacent to the AB are below El 109 ft and there is approximately 7 ft of concrete between the piping penetrating the containment in the penetration area, and the CR.

See Figure 3.

The closest ESF components are the charging pumps and the RHR Heat Exchangers located in the AB.

Charging Pumps (EI 73 ft)

There are 2 ft thick concrete floors at El 100 ft, El 115 ft, and El 140 ft, and a minimum of 1 ft-4 in. concrete at El 128ft for a total of 7 ft-4 in. of concrete 9

Enclosure PG&E Letter DCL-16-015 shielding provided by all the floors in the AB between the charging pumps and the CR.

RHR Heat Exchangers (HXs) (E/115 ft)

The closest post-LOCA accident source to the CR are the RHR HXs located east of the CR below floor El 115 ft; there is a minimum of 6 ft of concrete not including slant path (through a 2 ft floor at El 115 ft, a 2 ft wall of the AB that runs along row L, and a 2 ft floor at El 140 ft). Note that due to the location of the RHR HX in a narrow shielded cubicle, and the relative location of the CR above, the slant path through the auxiliary building wall is expected to be significant.

Based on the above, it is concluded that the ESF piping/components carrying recirculation sump fluids that are located in the vicinity of the CR are heavily shielded. Thus the dose contribution from the above sources will be negligible.

RWST Shine:

The calculated CR dose due to RWST direct shine is negligible due to distance (approximately 200ft) between the center-line of the RWST and the CR operator, and the intermediate shielding (note: the combined shielding provided by the East walls of the Fuel Handling Building (FHB) and the CR is 4ft which is the minimum shielding that can be credited). See Figure 12.

Technical Support Center As shown in Figure 1, the TSC is located on the southwest side of the plant at elevation 104 ft-4.5 in. in Plant ,Area "M" and is a significant distance away from the Containment and AB. The TSC is composed of four (4) rooms, i.e., a PG&E Office, an Operating Center, a Computation Center, and Records Management Room/NRC Office. See Figure 13 for the layout of the rooms.

Each room is approximately 29ft by 27 ft by 13.5 ft high. The rooms are side by side, separated by 26 in. thick concrete walls running north to south. The TSC has a ceiling of 12 in. of concrete. The north and south walls of the TSC are 26 in. thick concrete.

The east and west walls of the TSC are 16 in. and 18 in. of concrete, respectively.

The MER supporting the TSC is adjacent to the NRC office south wall. Two of the three outside entrances to the TSC are through the 16 in. thick concrete east wall and face the containment. All three entrances have a labyrinth design as shown on Figures 14a, 14b, and 14c.

It is noted that there is an additional doorway in the 18 in. thick concrete west wall leading to the lavatories; however, there is no access from the lavatory area to the TSC.

The 5 ft-4~ in. doorway has been replaced with a gypsum wall, which is not credited for shielding. See Figure 14d.

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Enclosure PG&E Letter DCL-16-015 Containment shine (i.e., due to airborne activity in containment) constitutes about eighteen (18) percent of the direct shine dose, which includes the streaming through the personnel hatch.

Sixty four (64) percent of the direct shine dose is from cloud shine (i.e., the airborne activity resulting from containment leakage, ESF system leakage via the plant vent, ESF system leakage via the containment penetration areas, the RHR pump seal failure and MEDT releases).

The remaining eighteen (18) percent is due to direct and scattered gamma radiation originating from the normal operation intake high efficiency particulate airborne (HEPA) filter, and the Mode 4 charcoal and HEPA emergency intake filters through the wall, and the radiation scatter through the 72 in. by 28 in. penetration located near the ceiling of the wall separating the Mechanicai/HVAC Room and NRC Office (see Figure 15). The scatter dose is conservatively estimated by dose albedo approach. A Monte-Carlo calculation using MCNP5 computer code confirms that the dose albedo approach is conservative in estimating the scatter dose. A mesh tally plot that depicts the dose in the source cubicle (Mechanical Room), through the 72 in. by 28 in. penetration, and in the receptor room (NRC office) is presented in Figure 16. The circles in Figure 16 represent the spherical tally cells.

Due to distance and shielding, the dose contribution in the TSC from post-LOCA radiation sources located in the AB, and the sump fluid in the RWST is negligible.

NRC ARCB-RAI-7 RG 1. 183 Appendix A, Section 3. 3 states:

The mixing rate attributed to natural convection between sprayed and unsprayed regions of the containment buifding, provided that adequate flow exists between these regions, is assumed to be two turnovers of the unsprayed regions per hour, unless other rates are justified.

In Enclosure Attachment 4 it states:

In accordance with RG 1. 183, Appendix A, Section 3. 3, prior to [containment fan cooler unit] CFCU initiation, the dose consequence model assumed a mixing rate attributable to natural convection between the sprayed and unsprayed regions of 2 turnovers of the unsprayed region per hour.

However, there was no discussion presented that demonstrates that adequate flow exists between the sprayed and unsprayed regions when the CFCUs are not in operation. Therefore, please provide a discussion that demonstrates that adequate flow exists between the sprayed and unsprayed regions when the CFCUs are not in operation.

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Enclosure PG&E Letter DCL-16-015 PG&E Response An assumed mixing rate of 2 turnovers of the unsprayed region/hour prior to initiation of the containment fan cool units (CFCUs) is deemed reasonable (and conservative) based on the following regulatory guidance, DCPP containment design features, and physical phenomena expected within containment immediately following a large break LOCA.

The time period of applicability of the mixing rate of 2 turnovers of the unsprayed region/hour is between accident initiation at t=O sees, and when the CFCUs become operational at t=86 sees .

• Regulatory guidance provided in Regulatory Position 3.1 of Appendix A of RG 1.183, Revision 0, indicates the acceptability of the assumption that the radioactivity released from the fuel is mixed instantaneously and homogeneously throughout the free air volume of the primary containment in PWRs unless internal compartments limit air exchange capability. Inherent in this assumption is the expectation of complete mixing in the containment during the initial blowdown stage of the accident when fission products are released from the core.

• The internal design of the DCPP Units 1 and 2 containment structures allows air to circulate freely. The volume above the operating floor, which comprises the majority of the containment net free volume, does not have significant barriers to obstruct mixing. Cubicles and compartments within the containment below the operating floor are provided with openings near the top as well as bottom to allow air circulation, e.g., major floor openings at El140 ft (operating floor) include grated hatches at each of the 4 reactor coolant pumps and adjacent to the equipment hatch, 2 open stairwells, and cubicles that connect spaces below and above the operating floor due to open roof/floor configurations such as the steam generator (SG) cubicles and Pressurizer cubicle.

The phenomenon that facilitates considerable mixing in the containment atmosphere in the initial stages of a LOCA is that related to the force of the LOCA blowdown break effluent and the resultant steam/gas injection source to containment atmosphere.

Injection flow with high momentum (called a "jet"), and injection flow with lower momentum (called a "plume") will be released from the break location and will force mixing of the containment atmosphere.

NRC ARCB-RAI-8 In Enclosure Attachment 4, Section 7.2.3.4, it states that the residual heat removal (RHR) pump seal failure resulting in a filtered release via the plant vent is DCPP's licensing basis with respect to the worst case passive single failure in the RHR system and is being retained as a release pathway for the AST LOCA dose consequence 12

Enclosure PG&E Letter DCL-16-015 analysis. The analysis provided appears to be consistent with RG 1. 183 Appendix A, Regulatory Positions section 5 with the exception of Regulatory Position 5. 2.

RG 1. 183 Appendix A, Regulatory Position 5.2 states that the engineered safety features (ESF) leakage should be taken as two times the sum of the simultaneous leakage from all components in the ESF recirculation systems above which the technical specification, or licensee commitments to item 111.0.1.1 of NUREG-0737 would require declaring such systems inoperable.

The RHR pump seal failure is considered to be ESF leakage. However, the analysis does not take into account two times the leakage in accordance with RG 1. 183 Appendix A, Regulatory Position 5.2.

Please provide an analysis that is consistent with RG 1.183 Appendix A, Regulatory Position 5.2 or provide a technical evaluation of the deviation from RG 1.183.

PG&E Response DCPP TS 5.5.2, "Primary Coolant Sources Outside Containment," represents a program that provides controls to minimize leakage from those portions of systems outside containment that could contain highly radioactive fluids during a serious transient or accident, to levels as low as practicable. The systems include portions of Recirculation Spray, Safety Injection, Chemical and Volume Control, RHR, RCS Sample, and Liquid and Gaseous Radwaste Treatment Systems. The program includes the following:

a. Preventive maintenance and periodic visual inspection requirements; and

b. Integrated leak test requirements for each system at refueling cycle intervals or less.

As part of the above program and commitment to NUREG-0737, the leakage from systems outside containment that could contain highly radioactive ESF fluids during or after an accident, is monitored and controlled in accordance with DCPP surveillance Procedure STP M-86. This procedure establishes the allowable leakage and leakage evaluation criteria. If leakage exceeds the total allowed leakage, the ECCS Post-LOCA recirculation path outside of containment is to be declared inoperable.

The RHR pump seal failure that is addressed in the LOCA dose consequence analysis consistent with the DCPP licensing basis as a radioactivity release pathway is separate from the ESF leakage. The RHR pump seal failure is covered by the DCPP licensing basis (UFSAR Section 3.1.1.1) as a Passive Failure, defined when applied to a fluid system as a break in the pressure boundary resulting in abnormal leakage not exceeding 50 gallons per minute (gpm) for 30 minutes. The licensing basis passive failure scenario is assumed to be an RHR pump seal failure occurring at 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> following a LOCA for a duration of 30 minutes.

13

Enclosure PG&E Letter DCL-16-015 Consistent with RG 1.183, Appendix A, Regulatory Position 5.2, and SRP 15.6.5, Appendix B, Section Ill, the leakage addressed in the LOCA dose consequence analysis supporting AST LAR 15-03, is based on the maximum expected operational leakage and is two times the sum of the assumed simultaneous allowable leakage from all components in the recirculation systems listed above, and excluding the RHR pump seal failure, at which the TS would require declaring such systems to be out of service.

The allowable leakage is assumed to occur throughout the accident, starting at the earliest time that the recirculation mode is initiated. The release due to the RHR pump seal failure is not considered a part of operational leakage and is therefore not subject to the multiplier of 2 applied to the ESF leakage rate. Rather, the RHR pump seal failure is considered an environmental release pathway resulting from a passive failure with a regulatory-based prescribed release rate and time frame (i.e., 50 gpm for 30 mins at t=24hrs post-LOCA, see Standard Review Plan (SRP) 15.6.5, Appendix B, Section Ill and UFSAR Section 3.1.1.1 for detail).

Thus, the contribution of the RHR pump seal failure to the dose consequences is separate from the expected operational ESF leakage that is controlled and monitored as discussed above, and therefore, is not subject to the multiplier of two that is applied to the ESF leakage rate. The application of the ESF leakage and the RHR pump passive failure as independent release paths in the DCPP dose analysis presented in the AST LAR 15-03 is consistent with RG 1.183 Appendix A, Regulatory Position 5.2, and SRP 15.6.5, Appendix B, Section Ill.

NRC ARCB-RAI-9 In Enclosure Attachment 4, Section 7.2.3.5, it states that as part of this application, DCPP is proposing to establish administrative acceptance criteria to ensure the total as-tested back leakage into the refueling water storage tank (RWST) from the containment recirculation sump is less than or equal to 1 gallon per minute (gpm). However, there is no further discussion about the administrative acceptance criteria. Furthermore, 10 CFR 50.36(c)(2)(ii) states:

A technical specification limiting condition for operation of a nuclear reactor must be established for each item meeting one or more of the following criteria:

(A) Criterion *1 ...

(B) Criterion 2. A process variable, design feature, or operating restriction that is an initial condition of a design basis accident or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

(C) Criterion 3 .. .

(0) Criterion 4 .. .

The 1 gpm back leakage into the RWST is an initial condition of the design basis loss of coolant accident radiological consequence that assumes the failure of a fission product 14

Enclosure PG&E Letter DCL-16-015 barrier. Please explain how the DCPP Technical Specifications meet 10 CFR 50.36(c)(2)(ii)(B) for the RWST back leakage.

PG&E Response The applicable TS for RWST leakage is the TS 5.5.2 "Primary Coolant Sources Outside Containment." The Primary Coolant Sources Outside Containment Program provides administrative controls to minimize leakage from those portions of systems outside containment that could contain highly radioactive fluids during a serious transient or accident to levels as low as practicable. The program includes periodic visual inspection requirements and integrated leak test requirements at refueling cycle intervals or less. As stated in Section 2.3, "Planned Procedural Updates," in the technical report "Implementation of Alternate Source Terms Summary of Dose Analyses and Results," Revision 1, contained in Attachment 2 to the Enclosure of PG&E Letter DCL-15-152, "Response to NRC Request for Additional Information Regarding License Amendment Request 15-03, 'Application of Alternative Source Term'," dated December 17, 2015, the ESF system leak testing procedures (that are part of the Boundary Leakage Program invoked by TS 5.5.2) will be updated to establish administrative acceptance criteria to ensure verification that the total as-tested back leakage into the RWST from the containment recirculation sump is less than or equal to 1 gpm.

NRC ARCB-RAI-10 In Enclosure Attachment 4, Section 7.2.3.6, it states that as part of this application, DCPP is proposing to establish administrative acceptance criteria to ensure the total as-tested flow hard piped to the miscellaneous equipment drain tank (MEDT) is less than 950 cubic centimeters per minute (cc/min) of ESF system leakage and 484 cclmin of non-radioactive fluid leakage. However, there is no further discussion about the administrative acceptance criteria. Furthermore, 10 CFR 50.36(c)(2)(ii) states:

A technical specification limiting condition for operation of a nuclear reactor must be established for each item meeting one or more of the following criteria:

(A) Criterion 1 ...

(B) Criterion 2. A process variable, design feature, or operating restriction that is an initial condition of a design basis accident or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

(C) Criterion 3 .. .

(0) Criterion 4 .. .

The 950 cclmin leakage into the MEDT is an initial condition of the design basis loss of coolant accident radiological consequence that assumes the failure of a fission product barrier. Please explain how DCPP Technical Specifications meet 10 CFR 50.36(c)(2)(ii)(B) for this parameter.

15

Enclosure PG&E Letter DCL-16-015

• PG&E Response The applicable TS for the flow for the MEDT is TS 5.5.2 "Primary Coolant Sources Outside Containment." As stated in Section 2.3, "Planned Procedural Updates," in the technical report "Implementation of Alternate Source Terms Summary of Dose Analyses and Results," Revision 1, the ESF system leak testing procedures (that are part of the Boundary Leakage Program invoked by TS 5.5.2) will be updated to establish administrative acceptance criteria to ensure that the total as-tested flow hard piped to the MEDT is less than 950 cc/min of ESF system leakage and 484 cc/min of nonradioactive fluid.

NRC ARCB-RAI-11 In Enclosure Attachment 4, Section 7.2.3.1, "Containment PressureNacuum Relief Line Release," it states:

It is conservatively assumed that 40% of release flashes and is instantaneously and homogeneously mixed in the containment atmosphere, and that the activity associated with the volatiles, i.e., 100% of the noble gases and 40% of the iodine in the reactor coolant, is available for release to the environment via this pathway.

RG 1. 183 Appendix A, Regulatory Position 3. 8 states that the purge release evaluation should assume that 100% of the radionuclide inventory in the reactor coolant system liquid is released to the containment at the initiation of the LOCA and that this inventory should be based on the TS reactor coolant system equilibrium activity. RG 1. 183 Appendix A, Regulatory Position 3. 8 does not make any statements about a reactor coolant system liquid flashing fraction. However, it is the NRC staff's position that 100%

of the radionuclide inventory in the reactor coolant system liquid is released to the containment and therefore, the flashing fraction would be 100% which is conservative and meets the intent of RG 1. 183 Appendix A, Regulatory Position 3. 8.

Please provide a containment pressure I vacuum relief line release analysis that is consistent with RG 1. 183 Appendix A, Regulatory Position 3. 8, as described above or explain how the flashing fraction of 40% is more conservative than a flashing fraction of

.100%.

PG&E Response As noted in Section 7.2.3.1 of Attachment 4 of AST LAR 15-03, and in accordance with RG 1.183, Revision 0, Appendix A, Regulatory Position 3.8, the LOCA dose consequence analysis assumes that 100 percent of the radionuclide inventory in the primary coolant (assumed to be at TS levels), is released to the containment at T= 0 hours0 days <br />0 hours <br />0 weeks <br />0 months <br />.

16

Enclosure PG&E Letter DCL-16-015 As noted by the Staff, Regulatory Position 3.8 does not make any statements about a reactor coolant system liquid flashing fraction. Since no specific guidance was provided in RG 1.183, for purposes of establishing dose consequences, PG&E utilizes a flash fraction commensurate with reactor coolant conditions at full power. The calculated flash fraction is approximately 38 percent. The dose consequence analysis supporting AST LAR 15-03 uses a flash fraction of 40 percent, which is conservative when compared to the calculated value.

NRC ARCB-RAI-12 In Enclosure Attachment 4, Section 7.1, ({Control Room Design I Operation I Transport Model," it describes the basic operation of the control room ventilation system (CRVS),

provides the Mode 4 parameter values assumed in the dose consequence analysis and references a December 2012 control room tracer gas test for the maximum unfiltered in leakage to the control room envelope (CRE). However, the Mode 4 parameter values do not seem to match the description of the CRVS operation provided in Section 7. 1.

One example is that the discussion states that in Mode 4 the pressurization flow at either intake is between 650 - 900 cubic feet per minute (cfm) and the air is taken from the less contaminated intake, however, in the Mode 4 parameter table it shows the pressurization flow of 650 - 900 cfm and also includes a filtered intake of 550 - BOO cfm. It is not clear to the NRC staff if this is another outside air intake or if this is part of the recirculation flow path, therefore, please provide a simplified diagram of the CRVS and explain in further detail the operation CRVS including the specific flow rates through the components as compared to the values assumed in the dose consequence analyses.

In addition, please provide a summary describing the December 2012 CRE test results and the test configurations.

PG&E Response During CRVS Mode 4, there is no outside air intake other than the pressurization flow and unfiltered CR in leakage. As stated in Section 7.1 of Attachment 4 of AST LAR 15-03, the Mode 4 pressurization flow at either intake is between 650- 900 cfm of which 100 cfm enters the CR unfiltered due to backdraft damper leakage. Thus the filtered portion of the pressurization flow is 550- 800 cfm.

Provided below is a simplified diagram of the CRVS when in Mode 4 operation. The CRVS Mode 4 parameter values assumed in the dose consequence analyses (provided in Section 7.1 of Attachment 4 of AST LAR 15-03) are also provided in the table below.

To provide a better understanding of the CRVS operation when in Mode 4, the referenced tabular data has been expanded to include the flow symbols used in the diagram below. (Note the location numbers are discussed below.)

17

Enclosure PG&E Letter DCL-16-015 DCPP CRVS Mode 4 Operation PF + UFI UFI BDD 7 PF-BDD PF -+---...J....__---i::: >--.

Control Room FR CRVS Mode 4 Min Flow Max Flow Location(Note 1) Mode 4 CR Parameters Symbol ( cfm} (Note 2) (cfm}

1 Pressurization Flow PF 650 900 2 Backdraft damper Leakage BOD 100 100 3 Filtered Intake PF-BDD 550 800 4 CF= FR + PF-1800 2200 Charcoal Filter Flow BOD 5 Filtered Recirc Flow FR 1250 1400 6 Unfiltered lnleakage UFI 70 70 7 CR Exhaust Flow (EF) PF + UFI 720 970 Note 1: Locations 1-5 are listed on the attached operations valve identification drawings (OVIDs) presented in Figures 17 and 18. Locations 6 & 7 are CR inleakage and exhaust flow (not CRVS flows).

Note 2: For a minimum flow example, Location 4 (1800) = location 1 - location 2

+location 5 (650-100+1250).

December 2012 CRE Test Results and Configurations In December 2012, CR Habitability tracer gas in leakage testing was performed at DCPP in accordance with DCPP Procedure STP M-57, "Control Room Ventilation System Tracer Gas Test." The flow values in the above table encompass the test results.

Location numbers in the table correspond to numbers on Attachment 1, Figures 17 and 18 (DCPP operating valve identification drawings (OVID's)) and identify the approximate locations of the parameters on the OVIDs. The OVIDs are also highlighted to identify 18

Enclosure PG&E Letter DCL-16-015 the minimum in-service CRVS equipment. This assumes a safety injection signal generated from Unit 1, which causes a train of CR ventilation to swap to Mode 4 on the Unit 1 side and a pressurization fan on Unit 2 side to start.

The amount of unfiltered in leakage into the pressurized CRE was determined using the constant injection method for each of the Fans (S99 and S96) supplying pressurization air. Based on the test results of each pressurization fan, two additional tests were performed. Testing was then performed with the S99 Fan and the TSC ventilation system in operation. The final test had the S99 Fan in operation, Unit 1 in Mode 3 and Unit 2 in Mode 4. The selected alignments were based on previous tests performed at DCPP.

A total of 4 tests were performed at DCPP over the course of 4 nights to determine the total in leakage into the CRE under various system operational modes. Summaries of these test results are shown in the table below.

Summary of Results of CR In leakage for the Tested Configurations lnleakage with Tracer Gas Phase. Train Configuration Outside Air (SCFM) 1 S99, Unit 2 Mode 4 Unit 1 Idle 32 +/- 5 2 S96, Unit 1 Mode 4 Unit 2 Idle 25 +/- 10 3 S99, Unit 2 Mode 4 Unit 1 Idle + TSC 23 +/- 7 4 S99, Unit 2 Mode 4 Unit 1 Mode 3 7+/-9 NRC ARCB-RAI-13 In Enclosure Attachment 4, Section 7.2. 7, "Technical Support Center Dose," it describes the basic operation of the TSC ventilation system and then states that it utilizes the CRVS. However, it is not clear to the NRC staff how the TSC ventilation system and the CRVS are combined in operation. Therefore, please provide a simplified diagram of the TSC ventilation system/CRVS and explain in further detail the operation TSC/CRVS including the specific flow rates through the components as compared to the values assumed in the dose consequence analyses.

PG&E Response Provided below are simplified diagrams of the TSC ventilation system during Mode 1 and Mode 4 operation. To provide a better understanding of the TSC ventilation system, tabular data that present the ventilation flows (and include the flow symbols used in the diagram), are also presented.

• 19

Enclosure PG&E Letter DCL-16-015 DCPP TSC - Mode 1 Operation

~~

EF FNI

.... UFI Mode 1 DCPP TSC - Mode 4 Operation EF UFI PF

--+------+~ ~ ~ ~ ~_____.

FR Mode4 Parameter S~mbol TSC Vent S~stem Value Filtered (HEPA only) Mode 1 Intake Rate FNI 500 cfm Filtered (Carbon + HEPA) Mode 4 Intake PF 500 cfm Rate Mode 4 Filtered (Carbon + HEPA)

FR 500 cfm Recirculation Rate Unfiltered lnleakage UFI 60 cfm Mode 1: FNI + UFI Environmental Exhaust Rate EF Mode 4: PF + UFI 20

Enclosure PG&E Letter DCL-16-015 Simplified diagrams of the TSC ventilation system in combination with the CRVS during Mode 4 operation for the period 0 to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> after a LOCA, and 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> to 30 days after the LOCA, are provided in Figures 19 and 20, respectively.

Figures 19 and 20 show that the Units 1 and 2 CR emergency pressurization air intakes also serve the TSC. Thus, the redundant PG&E Design Class I radiation monitors associated with CRVS Mode 4 operation and located at each pressurization air intake (which serve to automatically select the least contaminated inlet at the beginning of the accident, and manually select the least contaminated inlet during the course of the accident), also serve the TSC.

NRC ARCB-RAI-14 In the Enclosure Section 2. 1, "Proposed Changes to Current Licensing Basis," item 22 PG&E requests the following:

Credit the following existing administrative controls reflected in plant procedures.

These administrative controls ensure the [fuel handling building] FHB is maintained at a negative pressure relative to atmosphere during movement of irradiated fuel in the spent fuel pool, thus ensuring that the environmental releases occur via the Unit vent.

• The movable wall is in place and secured.

• No exit door from the FHB is propped open.

• At least one FHB ventilation system exhaust fan is running.

Attachment 4, Section 7. 3 presents the [fuel handling accident] FHA. Credit for the above administrative controls is taken to facilitate that the post-accident environmental release of radioactivity occurs via the plant vent.

In Enclosure Attachment 4, Section 7.3, it states that the radioactivity release pathways following a FHA in the FHB are established taking into consideration the following administrative controls:

• The movable wall is put in place and secured.

• No exit door is propped open.

• One fuel handling building ventilation system (FHBVS) exhaust fan is operating.

It further states that operation of the FHBVS with a minimum of one exhaust fan operating and all significant openings administratively closed will ensure negative pressure in the FHB which will result in post-accident environmental release of radioactivity occurring via the plant vent. However, TS 3. 7.13, "FHB VS," does not require one exhaust fan to be operating, the movable wall to be in place and secured, or require the exit to be closed.

21

Enclosure PG&E Letter DCL-16-015 10 CFR 50.36(c)(2)(ii) states:

A technical specification limiting condition for operation of a nuclear reactor must be established for each item meeting one or more of the following criteria:

(A) Criterion 1 ...

(B) Criterion 2. A process variable, design feature, or operating restriction that is an initial condition of a design basis accident or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

(C) Criterion 3 .. .

(D) Criterion 4 .. .

In addition RG 1. 183, Regulatory Position 5. 1. 2 states That credit may be taken for accident mitigation features that are classified as safety-related, are required to be operable by technical specifications ... "

The administrative controls discussed above are initial conditions of the design basis FHA in the FHB radiological consequence that assumes the failure of a fission product barrier. Please explain how DCPP Technical Specification 3. 7. 13 meets 10 CFR 50.36(c)(2)(ii)(B) for these administrative control conditions and meets RG 1. 183, Regulatory Position 5. 1. 2.

PG&E Response The administrative controls to ensure that the environmental releases occur via the unit vent are not required to be in TS 3. 7.13 because they do not fall within the criteria for mandatory inclusion in the TS in 10 CFR 50.36(c)(2)(ii). They are not needed to obviate the possibility that an abnormal situation or event will give rise to an immediate threat to the public health and safety. The NRC has previously approved relocation of similar administrative controls from the TS to PG&E-controlled procedures. The NRC approved 14 relocated TS from the TS to the PG&E ECGs and FSAR and associated PG&E-controlled procedures as part of the conversion to the improved TS in NRC License Amendments 135 for each unit dated May 28, 1999. The relocated TS and associated ECGs and FSARs were:

Relocated TS Title ECG/FSAR 3/4.1.2.1 Reactivity Control Systems - Boration ECG 8.5 Systems- Flow Path- Shutdown 3/4.1.2.3 Reactivity Control Systems - Charging ECG 8.6 Pump - Shutdown 3/4.1.2.4 Reactivity Control Systems - Charging ECG 8.7 Pumps - Operating 22

Enclosure PG&E Letter DCL-16-015 Relocated TS Title ECG/FSAR 3/4.1.2.5 Reactivity Control Systems - Borated ECG 8.8 Water Source - Shutdown 3/4.1.2.6 Reactivity Control Systems- Borated ECG 8.9 Water Sources - Operating 3/4.3.3.2 Movable lncore Detectors ECG 48.1 3/4.3.3.4 Meteorological Instrumentation ECG 40.1 3/4.9.3 Refueling Operations - Decay Time ECG 42.1 3/4.9.5 Refueling Operations - Communications ECG 42.2 3/4.9.6 Refueling Operations - Manipulator ECG 42.3 Crane and Auxiliary Hoist 3/4.9.7 Refueling Operations - Crane Travel - ECG 42.4 Fuel Handling Building 3/4.9.10.2 Refueling Operations -Water Level - ECG 42.5 Reactor Vessel - (Control Rods) 3/4.9.13 Spent Fuel Shipping Cask Movement FSAR 9.1 3/4.10.4 Special Test Exceptions - Position ECG 41.2 Indication System - Shutdown The NRC provided the following conclusion for the 14 relocated specifications:

The 14 relocated specifications from the CTS discussed above are not required to be in the ITS because they do not fall within the criteria for mandatory inclusion in the TS in 10 CFR 50.36(c)(2)(ii). They are not needed to obviate the possibility that an abnormal situation or event will give rise to an immediate threat to the public health and safety. In addition, the NRC staff finds that sufficient regulatory controls exist under the regulations cited above to maintain the effect of the provisions in these specifications. The NRC staff has concluded that appropriate controls have been established for all of the current specifications, information, and requirements that are being moved to an EGG and the FSAR.

The administrative controls to ensure the FHB is maintained at a negative pressure relative to atmosphere during movement of irradiated fuel in the spent fuel pool (SFP) will be included in the UFSAR as shown in Section 15.5.22.2.2, "Activity Release Transport Model," of Attachment 4 to the Enclosure of PG&E Letter DCL-15-152, dated December 17, 2015, which states in part:

The radioactivity release pathways following an FHA in the FHB are established taking into consideration the following Administration Controls:

23

Enclosure PG&E Letter DCL-16-015 During fuel movement in the FHB:

a. The movable wall is put in place and secured

b. No exit door is propped open

c. One FHBVS exhaust fan is operating (The supply fan flow (if operating) has been confirmed by design to have less flow than the exhaust fan)

These administrative controls are already included in existing Diablo Canyon Operating Procedure OP B-8D, "Refueling Prerequisites."

NRC ARCB-RAI-15 In Enclosure Attachment 4, Section 7.3 it has been determined that for the FHA in the FHB, the actual release rate lambda based on the FHBVS exhaust (i.e., 8. 7 hr-1) is larger than the release rate applicable to {(a 2-hr release" per regulatory guidance (i.e.,

3.45 hr-1). Thus the larger exhaust rate lambda associated with FHBVS operation plus the exhaust rate lambda for the 500 cfm outleakage is utilized in the analysis.

Please provide the technical basis for why this approach is conservative for control room doses.

PG&E Response Although both release rate lambdas (8.7 h( 1 and 3.45 h( 1) will release essentially all the airborne activity in the FHB in two hours, the release case with a larger release rate lambda will release activity faster than that with a smaller release rate lambda.

At the beginning of the accident (t = 0), the activity release rate (Ci/hr) for the 8.7 h( 1 release lambda case is 2.5 times greater than that for the 3.45 h( 1 release lambda case. This multiplier decreases as release goes on . At t = 10.6 min (0.176 hr, ln(8.7/3.45)/(8.7-3.45) = 0.176), the activity release rates for the two cases are the same. After 10.6 minutes, the activity release rate for the 8.7 h( 1 release lambda case is less than that for the 3.45 h( 1 release lambda case.

If the CR intake remained constant over the 2-hour release period, the resulting doses for the two cases would be the same. However, the DCPP CR intake switches from Mode 1 normal operation (4200 cfm plus or minus 10 percent unfiltered intake), to Mode 4 pressurization operation (550 cfm filtered intake, 70 cfm unfiltered inleakage, and 100 cfm unfiltered leakage via the backdraft dampers) after detection of contaminated air by the CR intake monitors. Thus a greater activity release rate at the earlier stage of the accident will result in a greater activity intake into the CR and consequently more conservative CR doses. A more detailed quantitative explanation is provided below:

24

Enclosure PG&E Letter DCL-16-015 Use of a release rate lambda of 8.7 h( 1 (vs. 3.45 h( 1) based on the FHB ventilation exhaust fans to calculate the CR doses following a FHA in the FHB impacts:

• the monitor response time,

• the activity intake into the CR, and

• the CR dose contribution due to the 500 cfm outleakage from the doors/openings of the fuel handling building.

As noted in LAR Enclosure, Attachment 4, Section 7.3, the instrument time constant for the DCPP CR Intake Monitors (1-RM-25, 1-RM-26, 2-RM-25, and 2-RM-26) is 2 seconds. The monitor response time for a design basis FHA (which will release a significant amount of activity to the atmosphere), is less than 1 second for both release rate lambdas (0.155 sec for the 8. 7 h( 1 release lambda, 0.42 sec for the 3.45 h(1 release lambda, based on equation presented in Attachment 4, Section 7.3). The analysis conservatively assumes a 20 second monitor response time. The delay time for switching to Mode 4 operation is 32 seconds, which includes the signal processing time (2 seconds) and the damper closure time (1 0 seconds).

• As explained earlier, at the beginning of the accident, the activity release rate for the 8.7 h( 1 release lambda case is 2.5 times greater than that for the 3.45 h( 1 release lambda case. Therefore, the CR activity intake during the 32 second Mode 1 operation is approximately 2.5 times greater for the 8. 7 h( 1 release lambda case. At t = 32 sec, the CR ventilation operation is switched to Mode 4 which results in less contaminated air entering the CR (i.e., due to the location of the pressurization intakes and associated smaller X/Qs, and a filter efficiency of 93 percent for elemental/organic iodine). The analysis also considers the dose contribution from (a) unfiltered inleakage, and (b) the pressurization intake backdraft damper bypass leakage, into the CR. From t = 32 sec to t = 10.6 min, the CR activity intake rate for the 8.7 hr1 lambda case remains greater than that for the 3.45 h( 1 lambda case. After 10.6 minutes, the CR activity intake rate for the

8. 7 h( 1 release lambda case is less than that for the 3.45 h( 1 release lambda case.

Because the atmospheric dispersion factors at the pressurization intakes are substantially smaller and the pressurization flow is filtered, the CR dose due to releases from the plant vent is dominated by a) the activity intake during the initial 32 second Mode 1 operation (4200 cfm plus or minus 10 percent unfiltered intake) and b) the 70 cfm unfiltered inleakage during the Mode 4 operation. The above two pathways result in 98.6 percent of the total dose from the plant vent release for the 8.7 h( 1 release lambda case.

Since the average X/Q value at the Unit 1 and 2 normal operation intakes and the X/Q value for unfiltered in leakage location are comparable, the ratio of activity intake during the initial 32 second verses that due to 70 cfm unfiltered in leakage is approximately

[4620 cfm x 8.7 h( 1 x (32/3600) hr] I [70 cfm x 1.0] = 5.1, for the 8.7 h( 1 release lambda case; and is approximately [4620 cfm x 3.45 h( 1 x (32/3600) hr] I [70 cfm x 1.0] = 2.0, for the 3.45 h( 1 release lambda case. The above comparison clearly demonstrates that 25

Enclosure PG&E Letter DCL-16-015 the activity intake during the initial 32 seconds of the accident is the most significant contributor to the total CR dose due to FHB ventilation release.

A smaller release rate lambda for the FHB ventilation release will increase slightly the CR dose contributed by the 500 cfm outleakage from the doors/openings because more activity is retained in the building and available for the outleakage release. However, the CR dose contribution from the 500 cfm outleakage (equivalent to 9.48E-2 h( 1 release rate lambda) constitutes only 3.8 percent of the total dose from both the plant vent and the FHB outleakage. This small dose contribution from the FHB outleakage is due to a release rate that is two orders of magnitude smaller than the plant vent release rate, and a X/Q value that is greater than that applicable to the plant vent release, by only a factor of 4. Thus, a small increase of this dose component will not change the conclusion that using the 8.7 h( 1 release rate lambda for the FHB ventilation release is conservative.

• Based on above, and given that the radiation monitor response time is essentially instantaneous in both cases, it is concluded that using the 8. 7 h( 1 release rate lambda for the FHB ventilation release is more conservative than using the 3.45 h( 1 release rate lambda in calculating the CR dose for a FHA in the FHB.

NRC ARCB-RAI-16 Appendix B of Regulatory Guide (RG) 1.183, Regulatory Position 1.1 states:

The number of fuel rods damaged during the accident should be based on a conservative analysis that considers the most limiting case. This analysis should consider parameters such as the weight of the dropped heavy load or the weight of a dropped fuel assembly (plus any attached handling grapples), the height of the drop, and the compression, torsion, and shear stresses on the irradiated fuel rods. Damage to adjacent fuel assemblies, if applicable (e.g., events over the reactor vessel), should be considered.

In Enclosure Attachment 4 Section 7. 3, it postulates that a spent fuel assembly is dropped during refueling in the spent fuel pool (SFP) located in the FHB, or in the reactor cavity located in containment and that all the rods in the dropped assembly are damaged.

The NRC staff reviewed the licensee's FHA analysis. The FHA activity is assumed to be released from (1) the damaged fuel via the SFP to the FHB, or (2) from the damaged fuel via the reactor cavity to the containment, from which it is assumed to be released to the environment within a two hour period as an unfiltered release via the plant vent or the containment equipment hatch, respectively. Credit was taken for the operation of one fuel handling building ventilation system exhaust fan, and the closure of the FHB exit doors. Credit was not taken for the fuel handling building ventilation system filtration. When evaluating the dose to operators in the control room, it was assumed 26

Enclosure PG&E Letter DCL-16-015 that the control room normal intake radiation monitors would initiate CRVS pressurization mode 4 at 22 seconds following the start of the event.

DCPP technical specification (TS) 3.3. 7, "Control Room Ventilation System (CRVS)

Actuation Instrumentation," states that the CRVS actuation instrumentation for each function in Table 3.3. 7-1 shall be operable, and is applicable according to Table 3.3. 7-1, which is during MODES 1, 2, 3, 4, 5, and 6, and during movement of recently irradiated fuel assemblies.

DCPP TS 3.3.8, "Fuel Handling Building Ventilation System (FHBVS) Actuation Instrumentation," states that the FHBVS actuation instrumentation for each function in Table 3. 3. 8-1 shall be operable, and is applicable according to Table 3. 3. 8-1, which is during movement of recently irradiated fuel assemblies in the fuel handling building.

DCPP TS 3. 7. 10, "Control Room Ventilation System (CRVS)," states that two CRVS trains shall be operable, and is applicable during MODES 1, 2, 3, 4, 5, and 6, and during movement of recently irradiated fuel assemblies.

DCPP TS 3.7.13, "Fuel Handling Building Ventilation System (FHBVS)," states that two FHBVS trains shall be operable and is applicable during movement of recently irradiated fuel assemblies in the fuel handling building.

DCPP TS 3. 7. 15, "Spent Fuel Pool Water Level," states that the spent fuel pool water level shall be ;::: 23 feet over the top of irradiated fuel assemblies seated in the storage racks and is applicable during movement of irradiated fuel assemblies in the spent fuel pool.

DCPP TS 3. 9. 7, "Refueling Cavity Water Level," states that the refueling cavity water level shall be maintained ;::: 23 feet above the top of reactor vessel flange and is applicable during movement of irradiate fuel assemblies within containment.

Given that the licensee voluntarily has requested a change to its licensing basis, please explain how the proposed revised fuel handling accident analysis meets or bounds Regulatory Guide 1. 183, Appendix B Regulatory Position 1. 1. Specifically, taking into account that the technical specification applicability does not:

1. Require CRVS actuation instrumentation or CRVS to be operable when in no mode during the movement of other radioactive loads (those other than recently irradiated fuel such as sources) in the SFP and other loads (such as fresh fuel) over irradiated fuel in the FHB or containment.

2. Require the FHBVS actuation instrumentation to be operable, a FHBVS exhaust fan to be running, or exit doors to be closed, or any of the equipment to be operable during movement of other radioactive loads (not defined as "recently" irradiated fuel) or loads (such as fresh fuel assemblies or sources, etc.) over irradiated fuel in the FHB.

27

Enclosure PG&E Letter DCL-16-015

3. Require the SFP or reactor cavity water level to be at least 23 feet during the movement of other radioactive loads (those other than irradiated fuel such as sources) and other loads (such as a fresh fuel) over irradiated fuel in the FHB or containment.

In addition, clarify how the revised analysis determines the most limiting case and how the fuel handling analysis shows that the limiting case is not the drop of a fuel assembly or object other than a recently irradiated fuel assembly.

PG&E Response The AST FHA accident analysis assumes the movement of an irradiated fuel assembly at the earliest permitted time of greater than or equal to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> from removal of the critical reactor core. The FHA assumption that the failure of all rods in one entire fuel assembly that was irradiated is consistent with the current limiting licensing basis limiting case that was approved by the NRC in License Amendments 8 (Unit 1) and 6 (Unit 2), dated May 30, 1986. The use of the AST methodology does not result in a change in the limiting case of one entire fuel assembly that was irradiated. In the safety evaluation for Amendments 8 and 6, the NRC concluded:

Regardless of how many spent fuel assemblies are stored in the pool, all fuel rods in one entire, freshly discharged fuel assembly were assumed to be breached, releasing their entire gap activity in accordance with conservative assumptions in Regulatory Guide 1.25. The assumption that all fuel rods in one assembly rupture is conservative because the kinetic energy available for causing damage to a fuel assembly dropped through water is fixed by the drop distance. The kinetic energy associated with the maximum drop height for a fuel handling accident is not considered sufficient to rupture the equivalent number of fuel rods of one assembly in both the dropped assembly and the impacted assembly. The accident evaluation conservatively assumes all rods in one entire assembly rupture and release their gap activity.

The movement of loads other than irradiated fuel, including control rods, is controlled through PG&E administrative controls to ensure that the equipment used to handle loads within the reactor cavity and SFP regions will function as designed and that the equipment has sufficient load capacity for handling loads. These administrative controls used to be contained in the TS; however, the NRC previously approved relocation of these administrative controls from the TS to PG&E-controlled procedures as part of the conversion to the improved TS in NRC License Amendments 135 for each unit dated May28, 1999. The NRC approval was based on the administrative requirements not falling within the criteria for mandatory inclusion in the TS in 10 CFR 50.36(c)(2)(ii).

Originally, the TS contained TS 3/4.9.6, "Manipulator Crane and Auxiliary Hoist," that contained requirements for (1) the manipulator crane, to ensure a minimum capacity of 3250 pounds and an overload cutoff limit of no more than 2700 pounds, and (2) 28

Enclosure PG&E Letter DCL-16-015 the auxiliary hoist, to ensure a minimum capacity of 700 pounds and a load indicator which is used to prevent lifting loads greater than 600 pounds. The NRC approved relocation of these original TS requirements to PG&E controlled ECGs as part of Amendments 135, and PG&E included the requirements in ECG 42.3, "Refueling Operations - Manipulator Crane." The NRC provided the following conclusion for the relocated TS 3/4.9.6:

Administrative controls exist to ensure that the equipment used to handle fuel within the reactor pressure vessel will function as designed and that the equipment has sufficient load capacity for handling fuel assemblies or control rods or both.

The refueling machine is designed with interlocks to provide overload limits to prevent damage to refueling equipment and fuel assemblies. These limits are not taken credit for to mitigate the consequences of a DBA; nor do these limits represent initial condition assumptions of an accident analysis. Although these limits represent operating restrictions and Criterion 2 of the Final Policy Statement and 10 CFR 50.36 includes operating restrictions, Criterion 2 applies only to those operating restrictions required to preclude analyzed accidents and transients.

Therefore, the Final Policy Statement and 10 CFR 50.36 criteria for including these requirements in the ITS are not met. An EGG is an acceptable licensee-controlled document for this information. This relocation is acceptable.

Originally, the TS contained TS 3/4.9.7, "Crane Travel* Spent Fuel Storage Areas,"

that contained requirements for the SFP crane that required that loads in excess of 2500 pounds not travel over fuel assemblies in the SFP (1) to limit the effect of a dropped load on the release of the radioactivity from a single irradiated fuel assembly (i.e., limit the effect to the gap activity), and (2) to assure that a dropped load does not distort the fuel in the SFP racks so as to cause criticality. The NRC approved relocation of these original TS requirements to PG&E controlled ECGs as part of Amendments 135, and PG&E included the requirements in ECG 42.4, "Refueling Operations - Crane Travel - Fuel Handling Building." The NRC provided the following conclusion for the relocated TS 3/4.9.7:

The administrative monitoring of loads moving over the fuel storage racks serves as a backup to the crane interlocks and physical stops. Although the specification supporls the maximum refueling accident assumption in the DBA, the crane travel limits are not monitored and controlled during operation; they are checked on a periodic basis to ensure their operability. Although this limit represents operating restrictions and Criterion 2 of the Final Policy Statement and 10 CFR 50.36 includes operating restrictions, Criterion 2 applies only to those operating restrictions required to preclude analyzed accidents and transients. This limit is not required to preclude analyzed accidents and transients. Therefore, the Final Policy Statement and 10 CFR 50.36 criteria for including these requirements in the ITS are not met. An EGG is an acceptable licensee-controlled document for this information. This relocation is acceptable.

29

Enclosure PG&E Letter DCL-16-015 Originally, the TS contained TS 3/4.9.1 0.2, "Reactor Vessel Water Level," that contained requirements on the reactor vessel water level during movement of control rods within the reactor vessel in Mode 6. The NRC approved relocation of these original TS requirements to PG&E controlled EGGs as part of Amendments 135, and PG&E included the requirements in ECG 42.5, "Refueling Operations -Water Level -

Reactor Vessel (Control Rods)." The NRC provided the following conclusion for the relocated TS 3/4.9.10.2:

The specification limits the minimum water over the irradiated fuel in the reactor core during the movement of controls rods within the reactor vessel while in Mode 6. Although this minimum water /eve/limit is being relocated to an EGG, there is the remaining requirement in ITS 3. 9. 7 (CTS 314. 9. 10. 1) that limits the minimum water level above the reactor core during movement of irradiated fuel assemblies within the containment. The limits in CTS 314. 9. 10.2 are not an initial condition of a DBA or transient. Although this limit represents operating restrictions and Criterion 2 of the Final Policy Statement and 10 CFR 50.36 includes operating restrictions, Criterion 2 applies only to those operating restrictions required to preclude analyzed accidents and transients. Therefore, the Final Policy Statement and 10 CFR 50.36 criteria for including these requirements in the ITS are not met.

An EGG is an acceptable licensee-controlled document for this information. This relocation is acceptable.

To account for the possibility that a fuel assembly that has decayed such that the CRVS will not actuate, an additional analysis case was performed for determining the dose to the CR operator for a FHA in the FHB and a FHA in Containment. The case assumed a delayed FHA at fuel offload or a FHA during reload occurs at a time when the fuel has decayed to such an extent that the radiation environment at the CR normal intake radiation monitors is just below the CRVS setpoint (CRVS Mode 4 is not initiated). This analysis case is discussed in PG&E Letter DCL-15-152, dated December 17, 2015, Enclosure Attachment 2, Section 7.3, "Fuel Handling Accident (FHA)." The analyses determined that the dose consequences of a DBA FHA (72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> of decay after reactor shutdown) bound those associated with the delayed FHA for both the FHA in the FHB and the FHA in the containment.

30

Enclosure Attachment 1 PG&E Letter DCL-16-015 Figure 1 General Plant Layout/Areas showing Control Room and TSC Location relative to U1/U2 Containments j I I 1. I I ~ i !r 1

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Enclosure Attachment 1 PG&E Letter DCL-16-015 Figure 6 Elevation View of MER, CR, CRVS Filters and Penetrations 0 I I

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Enclosure Attachment 1 PG&E Letter DCL-16-015 Figure 7 Penetrations along Column Line L (i.e., the wall separating Main CR and Mechanical Room) 18 I

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Penetrations after Col 18 are mirror image to those shown, noting that only one CRVS filter train is assumed to be operable.

MER- Part of Pressure boundary- no cloud source X

Note: The Centerline of each Penetration is located 17'-3" above the Control Room Floor 3

Enclosure Attachment 1 PG&E Letter DCL-16-015 Figure 8 Containment and Control Room Model Plan View 121 I G -+ ---~*

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Enclosure Attachment 1 PG&E Letter DCL-16-015 Figure 9 Containment SW-QADCGGP Model SPH 1 & 2

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Enclosure Attachment 1 PG&E Letter DCL-16-015 Figure 12 Gamma-Ray Trace Model for RWST Shine I .4~

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Enclosure Attachment 1 PG&E Letter DCL-16-015 Figure 14c - Labyrinths in West TSC wall Door BU 206

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Enclosure Attachment 1 PG&E Letter DCL-16-015 Figure 15 Mechanicai/HVAC Room adjacent to the NRC Office TSC- Mechanical/ HVAC Room Filter inlet ~

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Enclosure Attachment 1 PG&E Letter DCL-16-015 Figure 16 TSC Filter Scatter analysis Dose map of Mechanical Room and NRC Office plotted from the MCNPS mesh tally (mrem) 43 . 9 --t

---

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Enclosure Attachment 1 PG&E Letter DCL-16-015 FIGURE 17- CRVS Operating Valve Identification Drawings (OVID's) 161 62 2

Enclosure Attachment 1 PG&E Letter DCL-16-015 FIGURE 18- CRVS Operating Valve Identification Drawings (OVID's) 67

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Enclosure Attachment 1 PG&E Letter DCL-16-015 Figure 19 TSC/CR HVAC Alignment (TSC- 0-2 hrs Post-LOCA; CR- 44 sec- 2 hrs Post-LOCA)

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Enclosure Attachment 1 PG&E Letter DCL-16-015 Figure 20 TSC/CR HVAC Alignment 2 hrs - 30 days Post-LOCA

~ Infiltration Ill"'

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Pacific Gas ,and Ele,ctrtc Company James M. Welsch Diablo Canyon Power Plant Vice President, Nuclear Generation P.O. Box 56 Avila Beach, CA 93424 805.545.3242 E*Mail: JMWl @pge.com February 1, 2016 PG&E Letter DCL-16-015 U.S. Nuclear Regulatory Commission 10 CFR 50.90 ATTN: Document Control Desk Washington, DC 20555-0001 Docket No. 50-275, OL-DPR-80 Docket No. 50-323, OL-DPR-82 Diablo Canyon Units 1 and 2 Response to NRC Request for Additional Information Regarding License Amendment Request 15-03, "Application of Alternative Source Term"

References:

1. PG&E Letter DCL-15-069, License Amendment Request 15-03, "Application of Alternative Source Term," dated June 17, 2015 (ADAMS Accession No. ML15176A539)

2. PG&E Letter DCL-15-105, Supplement to License Amendment Request 15-03, "Application of Alternative Source Term," dated August 31, 2015 (ADAMS Accession No. ML15243A363)

3. E-mail from NRC Project Manager Siva P. Lingam, "Diablo Canyon 1 and 2 - Requests for Additional Information for License Amendment Request 15-03 to Adopt the Alternative Source Term per 10 CFR 50.67 (TAC Nos. MF6399 and MF6400),"dated December 2, 2015

Dear Commissioners and Staff:

License Amendment Request (LAR) 15-03, "Application of Alternative Source Term" was submitted by Pacific Gas and Electric (PG&E) Letter DCL-15-069 (Reference 1) and supplemented by PG&E Letter DCL-15-1 05 (Reference 2).

In Reference 3, the NRC Radiation Protection and Consequence Branch (ARCB) requested additional information required to complete the review of LAR 15-03.

PG&E's responses to the ARCB Staff's questions are provided in the Enclosure.

This information does not affect the results of the technical evaluation or the no significant hazards consideration determination previously transmitted in References 1 and 2.

A member of the STARS (Strategic Teaming and Resource Sharing) Alliance Callaway

• Diablo Canyon

• Palo Verde*

• Wolf Creek

I~

Document Control Desk PG&E Letter DCL-16-015 February 1, 2016

• Page 2 PG&E makes no regulatory commitments (as defined by NEt 99-04) in this letter.

This letter includes no revisions to existing regulatory commitments.

If you have any questions, or require additional information, please contact Mr. Hossein Hamzehee at (805) 545-4720.

I state under penalty of perjury that the foregoing is true and correct.

Executed on February 1, 2016.

Sincerely,

~~~

James M. Welsch Vice President, Nuclear Generation e 1d7/4418/50705089 Enclosure cc: Diablo Distribution cc/enc: Marc L. Dapas, NRC Region IV Administrator Siva P. Lingam, NRR Project Manager Gonzalo L. Perez, Branch Chief, California Department of Public Health Binesh K. Tharakan, Acting NRC Senior Resident Inspector A member of the STARS (Strategic Teaming and Resource Sharing) Alliance Callaway

• Diablo Canyon

• Palo Verde*

• Wolf Creek

Enclosure PG&E Letter DCL-16-015 PG&E Response to NRC Request for Additional Information (RAI) Regarding License Amendment Request 15-03, "Application of Alternative Source Term" NRC ARCB-RAI-1 In the Enclosure to the application dated June 17, 2015 it states:

Full implementation of [alternate source term] AST for DCPP Units 1 and 2 does not include revising ... NUREG-0737 responses associated with shielding and vital area access.

However, RG 1.183, Regulatory Position 4.3, "Other Dose Consequences," states that:

The guidance provided in Regulatory Positions 4.1 and 4.2 should be used, as applicable, in re-assessing the radiological analyses identified in Regulatory Position 1.3.1, such as those in NUREG-0737. Design envelope source terms provided in NUREG-0737 should be updated for consistency with the A ST. In general, radiation exposures to plant personnel identified in Regulatory Position 1.3.1 should be expressed in terms of [total effective dose equivalent]

effective TEDE.

In evaluating the submittal, the NRC staff could not determine how RG 1. 183, Regulatory Position 4. 3 has been assessed for DCPP. Please provide additional information describing how Regulatory Position 4. 3 has been assessed for DCPP.

PG&E Response In accordance with Regulatory Guide (RG) 1.183, Regulatory Position 4.3 and summarized below, Diablo Canyon Power Plant (DCPP) has assessed the effect of reassessing the radiological analyses performed in support of NUREG-0737 (and identified in Regulatory Position 1.3.1 as also affected by the accident source term) using the dose calculation methodology identified in Regulatory Positions 4.1 and 4.2 of RG 1.183.

Post-Accident Access Shielding (NUREG-0737. 11.8.2)

DCPP shielding study, which was developed in response to NUREG-0737, Item 11.8.2, was submitted to the NRC via PG&E Letter DCL-84-260, "Radiation Shielding Review,"

Revision 3, dated July 12, 1984. This report documented the estimated radiation exposure to plant personnel performing vital missions in support of accident mitigation and safe shutdown following a Loss-of-Coolant Accident (LOCA).

The source terms used in the referenced plant shielding study were based on traditional TID-14844 assumptions; specifically those outlined in NUREG-0737, Item 11.8.2. The nuclide release fractions specified in the Alternate Source Term (AST) methodology

Enclosure PG&E Letter DCL-16-015 outlined in RG 1.183 Revision 0, differ from those outlined in TID-14844/ NUREG-0737.

The difference in the release fractions has the potential to affect the dose rates in vital areas where piping containing post-LOCA sump fluid are located.

As discussed in NUREG-0933, Generic Issue 187, a study conducted by Sandia National Labs, "Evaluation of Radiological Consequences of Design Basis Accident at Operating Reactors Using the Revised Source Term," dated September 28, 1998, showed that exposure to containment atmosphere sources developed based on traditional source term methodology and AST methodology produced similar integrated doses. This report also showed that the integrated AST doses from exposure to post-LOCA sump fluid did not exceed those based on TID-14884 assumptions until42 days after an event at a pressurized water reactor (PWR).

Based on the above study it can be concluded that the differences in the release fractions associated with AST methodology would have little impact on the local dose rates during the 30-day post-LOCA mission time. Since the local dose rates are not expected to be significantly impacted by AST during the first 30 days following a LOCA, the conclusions of the shielding study with respect to operator exposure would not significantly change by expressing the mission doses in terms of total effective dose equivalent (TEDE).

The doses to the operator in the Control Room (CR) and Technical Support Center (TSC) were recalculated as part of the AST application and are presented in AST LAR 15-03 (Reference 1) in terms of TEDE.

Post-Accident Sampling Capability (NUREG-0737. 11.8.3)

Technical Specification (TS) Amendments 149 (Unit 1) and 149 (Unit 2) eliminated the requirements to have and maintain the Post-Accident Sampling System.

The post-accident sampling ability has been retained as a potential resource for data, to be used at the discretion of Chemistry personnel. As discussed above, should this ability be exercised, the previously estimated local dose rates near the sample panel are not expected to be significantly impacted by AST during the first 30 days following a LOCA, and the operator dose due to exposure would not significantly change by expressing the mission doses in terms of TEDE.

Accident Monitoring Instrumentation (NUREG-0737, II.F.1)

Post-accident monitoring instrumentation is available to measure various plant parameters. This instrumentation is designed to detect and remain operable under postulated design basis conditions. Although the more realistic AST source term would potentially involve a larger dose to equipment exposed to sump water over long periods of time, the conclusions of NUREG-0933, Generic Issue 187, which specifically addresses this issue, stated that there would be no discernable risk reduction associated with modifying the design basis for equipment qualification to adopt AST.

2

Enclosure PG&E Letter OCL-16-015 Therefore, the ability of the accident monitoring instrumentation to meet NUREG-0737 requirements is not impacted by the AST.

Leakage Control (NUREG-0737, 111.0.1.1)

Section 5.5.2 of the TS establishes a program to reduce leakage from primary coolant sources outside containment. As discussed in Section 2.3, Item 5, Attachment 4 of LAR 15-03, the engineered safety feature (ESF) system leak testing procedures that are invoked by TS 5.5.2 will be updated to reflect administrative acceptance criteria that ensure that the allowable leakage is at least a factor of two less than that assumed in the dose consequences analyses performed in support of AST implementation.

Emergency Response Facilities (NUREG-0737, III.A.1.2)

Emergency response facilities are shared by both Units 1 and 2. The TSC and the Operational Support Center (OSC) are co-located. The dose consequences in the TSC/OSC have been evaluated using AST methodology. The Emergency Operations Facility (EOF) is located outside of the 10 mile Emergency Planning Zone (EPZ).

Control Room Habitability (NUREG-0737, 111.0.3.4)

Radiological analyses were performed to meet the requirements of Item 111.0.3.4 of NUREG-0737. These analyses ensure that the CR operators are adequately protected against the effects of accidental release of radioactive gases and that the plant can be safely operated or shutdown under design basis conditions. As part of the AST implementation, the radiological analyses supporting CR habitability following design basis accidents were recalculated in terms of TEOE dose conversion factors and acceptance criteria.

NRC ARCB-RAI-2 In Enclosure Attachment 4, Section 7.2.6 it states:

To address the existing licensing basis, a TEDE dose is estimated for operator access to the control room. Because RG 1. 183 does not provide guidance on determining the egress and ingress to the control room following an accident, the same inputs used to estimate the current licensing basis values for access to the control room, along with the associated dose estimate presented in the UFSAR, are used to determine the TEDE dose estimate for ingress/egress.

In addition, Enclosure Table 1 and Enclosure Attachment 4 Table 8.1-1 state that the control room total effective dose equivalent presented for the loss of coolant accident represents the operator dose due to occupancy which is 3. 7 rem and that the value shown in parenthesis represents that portion of the total dose reported that is the contribution of direct shine from contained sources/external cloud which is 0. 7 rem.

Furthermore, they stated that the dose to the control room operator during routine 3

Enclosure PG&E Letter DCL-16-015 access for the 30 day duration of the accident is discussed in Enclosure Attachment 4, Section 7.2.6 and summarized in the text of Enclosure Attachment 4, section 8.0 which is 0. 037 rem.

10 CFR 50. 67, "Accident source term," (b)(2) states:

The NRC may issue the amendment only if the applicant's analysis demonstrates with reasonable assurance that:

(i) .. .

(ii) .. .

(iii) Adequate radiation protection is provided to permit access to and occupancy of the control room under accident conditions without personnel receiving radiation exposures in excess of 0. 05 Sv (5 rem) total effective dose equivalent (TEDE) for the duration of the accident.

In order to meet 10 CFR 50. 67, the radiation dose for accessing the control room must be evaluated. PG&E proposes to use the same inputs used to estimate the current licensing basis values for accessing the control room. However, the AST uses a different source term than that in the current licensing basis. Therefore, please provide an analysis of the radiation dose received from accessing the control room that reflects the new source term proposed with implementing the A ST.

PG&E Response As noted in Section 7.2.6 of Attachment 4 of LAR 15-03, the radiation dose received by the operator during the outbound and inbound excursions from the CR to the site boundary during the 30-day period post-LOCA is developed taking into consideration:

a) Use of the new source term proposed with implementation of AST b) The additional post-LOCA fission product release pathways assessed in LAR 15-03.

The assessment presented in Section 7.2.6 of Attachment 4 of LAR 15-03 takes into consideration the following:

a) Transit to and from the CR is expected after the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> following the accident by which time the airborne levels inside containment has reduced

• significantly due to the use of active fission product removal mechanisms such as containment sprays, and by radioactive decay, b) The operator is protected from radioactive ESF fluids by the shielding provided by the buildings that house such equipment, and c) Routine ingress/egress to the CR during the 30-day period following a LOCA falls into the mission dose category as discussed in NUREG-0737, II. B. 2.

NUREG-0737, Item II. B. 2, states that leakage of systems outside containment need not be considered as potential sources.

4

Enclosure PG&E Letter DCL-16-015 In accordance with DCPP original licensing basis, radiation exposures to personnel during egress and ingress (i.e., during routine access to the CR for the duration of the accident) could result from the following sources:

a) Airborne activity in the containment leakage plume b) Direct gamma radiation from fission products in the containment structure As noted in Section 7 .2 .6 of Attachment 4 of LAR 15-03, no additional sources need to be addressed to assess the operator dose due to ingress/egress to the CR since all of the additional post-LOCA fission product release pathways assessed with this application fall either into the category of (a) are terminated prior to 24hr after accident initiation (i.e., containment vacuum relief and RHR pump seal failure), or (b) "releases due to leakage of systems outside containment," and therefore need not be addressed per the guidance provided in NUREG-0737, II. B. 2 (i.e., refueling water storage tank (RWST) and miscellaneous equipment drain tank (MEDT) releases).

Since the configuration and shielding associated with the containment structure and CR has not changed with this application, the only change is in the source term.

The effect of the change in source term on the original licensing basis dose to CR personnel during egress/ingress reported in the DCPP Updated Final Safety Analysis Report (UFSAR), Table 15.5-33, and the subsequent development of the ultimate TEDE dose, is addressed using scaling techniques, and discussed in detail in Section 7.2.6 of Attachment 4 of LAR 15-03.

NRC ARCB-RAI-3 Please provide the RAD TRAD input and output files, in electronic format, for each of the AST DBAs described in the LAR.

PG&E Response In response to PG&E request, and as agreed to in the e-mail from Siva Lingam (NRC) to K. Schrader (PG&E) dated December 29, 2015, the RADTRAD input and output files were provided for NRC review during the January 12-14, 2016, NRC Audit of the calculations supporting DCPP AST LAR 15-03. This approach is in lieu of submitting the RADTRAD input and output files (considered to be proprietary by the vendor) with a request for withholding from the public via this response.

NRC ARCB-RAI-4 Please confirm that the failed fuel percentage (1 0%) stated in the Enclosure Attachment 4 Table 4.3-1 note, which is applied to the control rod ejection accident and the locked rotor accident, is equivalent to 10% of the rods in the core as opposed to 10% of the rods in an assembly.

5

Enclosure PG&E Letter DCL-16-015 PG&E Response PG&E confirms that the failed fuel percentage ( 10 percent) stated in the Enclosure Table 4.3-1 note, which is applied to the control rod ejection accident and the locked rotor accident, is equivalent to 10 percent of the rods in the core.

NRC ARCB-RAI-5 DELETED.

NRC ARCB-RAI-6 Enclosure Attachment 4, Section 6. 2, ((Direct Shine Dose from External and Contained Sources," states:

CB&I S&W Inc. [Chicago Bridge and Iron Stone and Webster Incorporated] [A CB&I Company] point kernel shielding computer program SW-QADCGGP is used to calculate the deep dose equivalent (DOE) in the control room, [technical support center] TSC and at the [exclusion area boundary] EAB due to external and contained sources. The calculated DOE is added to the inhalation (CEDE) and the submersion factors are used and the geometry models are prepared to ensure that un-accounted streaming/scattering paths were eliminated. The dose albedo method with conservative albedo values is used to estimate the scatter dose in situations where the scattering contributions are potentially significant.

[American National Standards Institute/American Nuclear Society] ANSIIANS 6.1.1-1977 ((neutron and gamma-ray flux-to-dose-rate factors" (Reference 31) is used to convert the gamma flux to the dose equivalent rate.

Enclosure Attachment 4, Section 7.2. 7, ((Technical Support Center Dose," provides further detail on the direct shine dose to the technical support center from external and contained sources. Enclosure Attachment 4, Section 7. 2. 7, states:

CB&I S&W Inc. computer code PERC2 is used to calculate the dose to TSC personnel due to airborne radioactivity releases following a [loss of coolant accident] LOCA. The direct shine dose to an operator in the TSC due to contained or external sources resulting from a postulated LOCA is calculated using CB&I S&W Inc. point kernel shielding computer program SW-QADCGGP.

The post-LOCA gamma energy release rated [mega electron volt per second]

(Me Vlsec) and integrated gamma energy release [mega electron volt hour per second] (Me V-hrlsec) in the various external sources are developed with computer program PERC2.

In evaluating the LAR, the NRC staff could not thoroughly review and perform confirmatory calculations of the direct shine dose due to external and contained sources at the control room, TSC, low population zone (LPZ), and at the EAB from the 6

Enclosure PG&E Letter DCL-16-015 discussions described in the LAR. Therefore, please provide additional information in enough detail that will enable the NRC staff to be able to perform an independent calculation of the direct shine dose to the control room and the TSC from external and contained sources.

PG&E Response All Figures referenced in this response are included in enclosed Attachment 1. Figure 1 provides a plant layout and the general location of the CR and the TSC.

Figures 2 through 12 provide the layout arrangement of the CR, the location of radiation sources with respect to the CR, and pictorial representations of some of the shielding models used to simulate the plant configuration.

Figures 13 through 16 provide similar information for the TSC.

Control Room The CR is located in Plant Area "H" as shown on Figure 1. The overall plan view (Elevation (EI) 140ft) of the Unit 1 and Unit 2 shared CR is illustrated in Figure 2. The numbers in the rectangular boxes are plant room ID numbers and the numbers in the diamond boxes are plant door ID numbers. The overall dimension of the main CR is 118 ft by 70ft, and the room height is 20 ft. The Mechanical Equipment Room (MER) that serves the main CR is located at El154 ft-6 inch (in.) along column line L. The MER is included in the CR emergency ventilation system envelope. See Figure 6.

The post-LOCA sources addressed herein for direct shine include:

a) Airborne activity in containment b) Buildup of activity on the control room ventilation system (CRVS) filters due to accumulation of activity resulting from 5 airborne release pathways, i.e.,

containment leakage, ESF system leakage via the plant vent, ESF system leakage via the containment penetration areas, the residual heat removal (RHR) pump seal failure and MEDT releases. RWST releases are minimal compared to the MEDT releases; therefore, the activity accumulation in the CRVS filter due to RWST release is not addressed.

c) The activity concentration in the contaminated cloud outside the CR pressure boundary due to environmental releases via the 5 release pathways discussed above, i.e., containment leakage, ESF system leakage via the plant vent, ESF system leakage via the containment penetration areas, the RHR pump seal failure, and MEDT releases. Due to the reason outlined above for the CRVS filter, the cloud associated with the RWST release is not addressed.

7

Enclosure PG&E Letter DCL-16-015 d) Recirculating sump water activity in piping and equipment.

e) Sump water activity in the RWST due to RWST back leakage.

Containment Shine:

The calculated dose from direct shine of activity airborne in the containment is about 0.6 percent of the total direct shine DOE dose. This dose contribution is small because (a) the CR outer wall facing the containment is made of 3 ft thick concrete, the roof of the CR is made of 3ft 4 in. thick concrete and there is significant slant path through the remaining 2ft thick concrete CR walls and floor (due to the geometry). The cylindrical wall of the containment is 3 ft-8 in. thick concrete; (b) the azimuth of personnel hatch that is located at the same elevation as the CR is such that there is negligible contribution to the direct shine dose in the CR; and (c) other large containment penetrations either don't face the CR or, as in the instance of a Main Steam (MS) line penetration, are below the CR elevation. Both the personnel hatch and MS line penetration are modeled using QAD-CGGP. For an understanding of the configuration, see Figure 3 (elevation view), Figure 8 (plan view of containment and CR, model has an axis shift of 180 degrees for modelling convenience), Figure 9 (QAD bulk shielding containment model), and Figure 10 (shows the QAD model of the 2ft thick concrete shield wall in front of the 115 in~ diameter personnel hatch).

CR Filter Shine:

The calculated dose from direct and scattered radiation due to activity buildup on the CR emergency filters is small, i.e., about 0.6 percent of the total direct shine dose.

This is due to several reasons (a) the filter is located in the MER about 23ft beyond column L, (b) the wall separating the MER and CR is 2ft concrete, (c) the penetrations in this wall along column L are 17 ft-3 in. above the CR floor, and (d) the largest of the six penetrations in the wall is 30 in. by 26 in. See Figures 6 and 7.

Cloud Shine:

Contaminated cloud shine through doors in the outer walls of the CR is the most significant direct shine dose contributor to personnel located in the CR; especially through doors 509 and 508 at column line H (see Figure 2). Airborne radioactivity concentrations in the stairwell between doors 508 and 509 are assumed to be the same as the concentrations in the environment outside resulting from all identified leakages (i.e., containment, ESF, MEDT and the passive component failure). The radioactivity concentrations in the external cloud are based on atmospheric dispersion factors applicable to a receptor located at the center of the CR boundary at roof level. Credit is taken for solid angle reduction through the door openings. Since doors 508 and 509 are adjacent to the Turbine Building (TB), no credit is taken for shielding offered by the TB (mostly corrugated sheet metal). The doorway along column L leads into the Auxiliary Building (AB) and concrete shielding in the AB is credited.

8

Enclosure PG&E Letter DCL-16-015 Currently there is a large opening in the wall along column H that leads into Room 507 (briefing room), however, as noted in Section 2.2 of Attachment 4 and Commitment 1 in of LAR 15-03, prior toAST implementation, PG&E will install shielding material at this location equivalent to that provided by the CR outer walls. Currently, administrative controls are in place to restrict post-LOCA access to Room 507.

There are a significant number of small floor penetrations in the CR ranging from 4 in. by 4 in. to 12 in. by 16 in. or 8 in. by 20 in. The penetrations are mostly small and spread out so they do not combine to make an "effectively" large penetration. CR panels are located on top of most (if not all) of these cable penetrations, so standing on top of a penetration is unlikely. Two of the largest penetrations are modeled, i.e.,

a 12 in. by 16 in. and an 8 in. by 20 in. penetration. The CR operator direct and scattered dose from airborne activity in the cable spreading room below the CR is conservatively based on the bounding penetration centerline dose. About 15 percent of the total direct shine dose comes from these floor penetrations.

ESF Component Shine:

The direct shine dose from pipes and components carrying recirculating containment sump water is negligible. Listed below is the relevant information associated with the location and thickness of concrete shielding available between the CR and the piping I components carrying ESF fluids that are located in the vicinity of the CR.

As shown in Figure 3 and discussed below, the CR which is located in the AB has significant shielding between it and the closest ESF components that are carrying post-accident radioactive fluids.

Control Room - El140ft Cable Spreading Room - El128 ft Battery Room - El 115ft Switchgear Room - El1 00 ft ChernE Office I Lab- El 85ft CCW I Charging pumps - El 73ft Laundry- El 60ft The pipe penetrations in the containment wall in the containment penetration area adjacent to the AB are below El 109 ft and there is approximately 7 ft of concrete between the piping penetrating the containment in the penetration area, and the CR.

See Figure 3.

The closest ESF components are the charging pumps and the RHR Heat Exchangers located in the AB.

Charging Pumps (EI 73 ft)

There are 2 ft thick concrete floors at El 100 ft, El 115 ft, and El 140 ft, and a minimum of 1 ft-4 in. concrete at El 128ft for a total of 7 ft-4 in. of concrete 9

Enclosure PG&E Letter DCL-16-015 shielding provided by all the floors in the AB between the charging pumps and the CR.

RHR Heat Exchangers (HXs) (E/115 ft)

The closest post-LOCA accident source to the CR are the RHR HXs located east of the CR below floor El 115 ft; there is a minimum of 6 ft of concrete not including slant path (through a 2 ft floor at El 115 ft, a 2 ft wall of the AB that runs along row L, and a 2 ft floor at El 140 ft). Note that due to the location of the RHR HX in a narrow shielded cubicle, and the relative location of the CR above, the slant path through the auxiliary building wall is expected to be significant.

Based on the above, it is concluded that the ESF piping/components carrying recirculation sump fluids that are located in the vicinity of the CR are heavily shielded. Thus the dose contribution from the above sources will be negligible.

RWST Shine:

The calculated CR dose due to RWST direct shine is negligible due to distance (approximately 200ft) between the center-line of the RWST and the CR operator, and the intermediate shielding (note: the combined shielding provided by the East walls of the Fuel Handling Building (FHB) and the CR is 4ft which is the minimum shielding that can be credited). See Figure 12.

Technical Support Center As shown in Figure 1, the TSC is located on the southwest side of the plant at elevation 104 ft-4.5 in. in Plant ,Area "M" and is a significant distance away from the Containment and AB. The TSC is composed of four (4) rooms, i.e., a PG&E Office, an Operating Center, a Computation Center, and Records Management Room/NRC Office. See Figure 13 for the layout of the rooms.

Each room is approximately 29ft by 27 ft by 13.5 ft high. The rooms are side by side, separated by 26 in. thick concrete walls running north to south. The TSC has a ceiling of 12 in. of concrete. The north and south walls of the TSC are 26 in. thick concrete.

The east and west walls of the TSC are 16 in. and 18 in. of concrete, respectively.

The MER supporting the TSC is adjacent to the NRC office south wall. Two of the three outside entrances to the TSC are through the 16 in. thick concrete east wall and face the containment. All three entrances have a labyrinth design as shown on Figures 14a, 14b, and 14c.

It is noted that there is an additional doorway in the 18 in. thick concrete west wall leading to the lavatories; however, there is no access from the lavatory area to the TSC.

The 5 ft-4~ in. doorway has been replaced with a gypsum wall, which is not credited for shielding. See Figure 14d.

10

Enclosure PG&E Letter DCL-16-015 Containment shine (i.e., due to airborne activity in containment) constitutes about eighteen (18) percent of the direct shine dose, which includes the streaming through the personnel hatch.

Sixty four (64) percent of the direct shine dose is from cloud shine (i.e., the airborne activity resulting from containment leakage, ESF system leakage via the plant vent, ESF system leakage via the containment penetration areas, the RHR pump seal failure and MEDT releases).

The remaining eighteen (18) percent is due to direct and scattered gamma radiation originating from the normal operation intake high efficiency particulate airborne (HEPA) filter, and the Mode 4 charcoal and HEPA emergency intake filters through the wall, and the radiation scatter through the 72 in. by 28 in. penetration located near the ceiling of the wall separating the Mechanicai/HVAC Room and NRC Office (see Figure 15). The scatter dose is conservatively estimated by dose albedo approach. A Monte-Carlo calculation using MCNP5 computer code confirms that the dose albedo approach is conservative in estimating the scatter dose. A mesh tally plot that depicts the dose in the source cubicle (Mechanical Room), through the 72 in. by 28 in. penetration, and in the receptor room (NRC office) is presented in Figure 16. The circles in Figure 16 represent the spherical tally cells.

Due to distance and shielding, the dose contribution in the TSC from post-LOCA radiation sources located in the AB, and the sump fluid in the RWST is negligible.

NRC ARCB-RAI-7 RG 1. 183 Appendix A, Section 3. 3 states:

The mixing rate attributed to natural convection between sprayed and unsprayed regions of the containment buifding, provided that adequate flow exists between these regions, is assumed to be two turnovers of the unsprayed regions per hour, unless other rates are justified.

In Enclosure Attachment 4 it states:

In accordance with RG 1. 183, Appendix A, Section 3. 3, prior to [containment fan cooler unit] CFCU initiation, the dose consequence model assumed a mixing rate attributable to natural convection between the sprayed and unsprayed regions of 2 turnovers of the unsprayed region per hour.

However, there was no discussion presented that demonstrates that adequate flow exists between the sprayed and unsprayed regions when the CFCUs are not in operation. Therefore, please provide a discussion that demonstrates that adequate flow exists between the sprayed and unsprayed regions when the CFCUs are not in operation.

11

Enclosure PG&E Letter DCL-16-015 PG&E Response An assumed mixing rate of 2 turnovers of the unsprayed region/hour prior to initiation of the containment fan cool units (CFCUs) is deemed reasonable (and conservative) based on the following regulatory guidance, DCPP containment design features, and physical phenomena expected within containment immediately following a large break LOCA.

The time period of applicability of the mixing rate of 2 turnovers of the unsprayed region/hour is between accident initiation at t=O sees, and when the CFCUs become operational at t=86 sees .

• Regulatory guidance provided in Regulatory Position 3.1 of Appendix A of RG 1.183, Revision 0, indicates the acceptability of the assumption that the radioactivity released from the fuel is mixed instantaneously and homogeneously throughout the free air volume of the primary containment in PWRs unless internal compartments limit air exchange capability. Inherent in this assumption is the expectation of complete mixing in the containment during the initial blowdown stage of the accident when fission products are released from the core.

• The internal design of the DCPP Units 1 and 2 containment structures allows air to circulate freely. The volume above the operating floor, which comprises the majority of the containment net free volume, does not have significant barriers to obstruct mixing. Cubicles and compartments within the containment below the operating floor are provided with openings near the top as well as bottom to allow air circulation, e.g., major floor openings at El140 ft (operating floor) include grated hatches at each of the 4 reactor coolant pumps and adjacent to the equipment hatch, 2 open stairwells, and cubicles that connect spaces below and above the operating floor due to open roof/floor configurations such as the steam generator (SG) cubicles and Pressurizer cubicle.

The phenomenon that facilitates considerable mixing in the containment atmosphere in the initial stages of a LOCA is that related to the force of the LOCA blowdown break effluent and the resultant steam/gas injection source to containment atmosphere.

Injection flow with high momentum (called a "jet"), and injection flow with lower momentum (called a "plume") will be released from the break location and will force mixing of the containment atmosphere.

NRC ARCB-RAI-8 In Enclosure Attachment 4, Section 7.2.3.4, it states that the residual heat removal (RHR) pump seal failure resulting in a filtered release via the plant vent is DCPP's licensing basis with respect to the worst case passive single failure in the RHR system and is being retained as a release pathway for the AST LOCA dose consequence 12

Enclosure PG&E Letter DCL-16-015 analysis. The analysis provided appears to be consistent with RG 1. 183 Appendix A, Regulatory Positions section 5 with the exception of Regulatory Position 5. 2.

RG 1. 183 Appendix A, Regulatory Position 5.2 states that the engineered safety features (ESF) leakage should be taken as two times the sum of the simultaneous leakage from all components in the ESF recirculation systems above which the technical specification, or licensee commitments to item 111.0.1.1 of NUREG-0737 would require declaring such systems inoperable.

The RHR pump seal failure is considered to be ESF leakage. However, the analysis does not take into account two times the leakage in accordance with RG 1. 183 Appendix A, Regulatory Position 5.2.

Please provide an analysis that is consistent with RG 1.183 Appendix A, Regulatory Position 5.2 or provide a technical evaluation of the deviation from RG 1.183.

PG&E Response DCPP TS 5.5.2, "Primary Coolant Sources Outside Containment," represents a program that provides controls to minimize leakage from those portions of systems outside containment that could contain highly radioactive fluids during a serious transient or accident, to levels as low as practicable. The systems include portions of Recirculation Spray, Safety Injection, Chemical and Volume Control, RHR, RCS Sample, and Liquid and Gaseous Radwaste Treatment Systems. The program includes the following:

a. Preventive maintenance and periodic visual inspection requirements; and

b. Integrated leak test requirements for each system at refueling cycle intervals or less.

As part of the above program and commitment to NUREG-0737, the leakage from systems outside containment that could contain highly radioactive ESF fluids during or after an accident, is monitored and controlled in accordance with DCPP surveillance Procedure STP M-86. This procedure establishes the allowable leakage and leakage evaluation criteria. If leakage exceeds the total allowed leakage, the ECCS Post-LOCA recirculation path outside of containment is to be declared inoperable.

The RHR pump seal failure that is addressed in the LOCA dose consequence analysis consistent with the DCPP licensing basis as a radioactivity release pathway is separate from the ESF leakage. The RHR pump seal failure is covered by the DCPP licensing basis (UFSAR Section 3.1.1.1) as a Passive Failure, defined when applied to a fluid system as a break in the pressure boundary resulting in abnormal leakage not exceeding 50 gallons per minute (gpm) for 30 minutes. The licensing basis passive failure scenario is assumed to be an RHR pump seal failure occurring at 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> following a LOCA for a duration of 30 minutes.

13

Enclosure PG&E Letter DCL-16-015 Consistent with RG 1.183, Appendix A, Regulatory Position 5.2, and SRP 15.6.5, Appendix B, Section Ill, the leakage addressed in the LOCA dose consequence analysis supporting AST LAR 15-03, is based on the maximum expected operational leakage and is two times the sum of the assumed simultaneous allowable leakage from all components in the recirculation systems listed above, and excluding the RHR pump seal failure, at which the TS would require declaring such systems to be out of service.

The allowable leakage is assumed to occur throughout the accident, starting at the earliest time that the recirculation mode is initiated. The release due to the RHR pump seal failure is not considered a part of operational leakage and is therefore not subject to the multiplier of 2 applied to the ESF leakage rate. Rather, the RHR pump seal failure is considered an environmental release pathway resulting from a passive failure with a regulatory-based prescribed release rate and time frame (i.e., 50 gpm for 30 mins at t=24hrs post-LOCA, see Standard Review Plan (SRP) 15.6.5, Appendix B, Section Ill and UFSAR Section 3.1.1.1 for detail).

Thus, the contribution of the RHR pump seal failure to the dose consequences is separate from the expected operational ESF leakage that is controlled and monitored as discussed above, and therefore, is not subject to the multiplier of two that is applied to the ESF leakage rate. The application of the ESF leakage and the RHR pump passive failure as independent release paths in the DCPP dose analysis presented in the AST LAR 15-03 is consistent with RG 1.183 Appendix A, Regulatory Position 5.2, and SRP 15.6.5, Appendix B, Section Ill.

NRC ARCB-RAI-9 In Enclosure Attachment 4, Section 7.2.3.5, it states that as part of this application, DCPP is proposing to establish administrative acceptance criteria to ensure the total as-tested back leakage into the refueling water storage tank (RWST) from the containment recirculation sump is less than or equal to 1 gallon per minute (gpm). However, there is no further discussion about the administrative acceptance criteria. Furthermore, 10 CFR 50.36(c)(2)(ii) states:

A technical specification limiting condition for operation of a nuclear reactor must be established for each item meeting one or more of the following criteria:

(A) Criterion *1 ...

(B) Criterion 2. A process variable, design feature, or operating restriction that is an initial condition of a design basis accident or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

(C) Criterion 3 .. .

(0) Criterion 4 .. .

The 1 gpm back leakage into the RWST is an initial condition of the design basis loss of coolant accident radiological consequence that assumes the failure of a fission product 14

Enclosure PG&E Letter DCL-16-015 barrier. Please explain how the DCPP Technical Specifications meet 10 CFR 50.36(c)(2)(ii)(B) for the RWST back leakage.

PG&E Response The applicable TS for RWST leakage is the TS 5.5.2 "Primary Coolant Sources Outside Containment." The Primary Coolant Sources Outside Containment Program provides administrative controls to minimize leakage from those portions of systems outside containment that could contain highly radioactive fluids during a serious transient or accident to levels as low as practicable. The program includes periodic visual inspection requirements and integrated leak test requirements at refueling cycle intervals or less. As stated in Section 2.3, "Planned Procedural Updates," in the technical report "Implementation of Alternate Source Terms Summary of Dose Analyses and Results," Revision 1, contained in Attachment 2 to the Enclosure of PG&E Letter DCL-15-152, "Response to NRC Request for Additional Information Regarding License Amendment Request 15-03, 'Application of Alternative Source Term'," dated December 17, 2015, the ESF system leak testing procedures (that are part of the Boundary Leakage Program invoked by TS 5.5.2) will be updated to establish administrative acceptance criteria to ensure verification that the total as-tested back leakage into the RWST from the containment recirculation sump is less than or equal to 1 gpm.

NRC ARCB-RAI-10 In Enclosure Attachment 4, Section 7.2.3.6, it states that as part of this application, DCPP is proposing to establish administrative acceptance criteria to ensure the total as-tested flow hard piped to the miscellaneous equipment drain tank (MEDT) is less than 950 cubic centimeters per minute (cc/min) of ESF system leakage and 484 cclmin of non-radioactive fluid leakage. However, there is no further discussion about the administrative acceptance criteria. Furthermore, 10 CFR 50.36(c)(2)(ii) states:

A technical specification limiting condition for operation of a nuclear reactor must be established for each item meeting one or more of the following criteria:

(A) Criterion 1 ...

(B) Criterion 2. A process variable, design feature, or operating restriction that is an initial condition of a design basis accident or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

(C) Criterion 3 .. .

(0) Criterion 4 .. .

The 950 cclmin leakage into the MEDT is an initial condition of the design basis loss of coolant accident radiological consequence that assumes the failure of a fission product barrier. Please explain how DCPP Technical Specifications meet 10 CFR 50.36(c)(2)(ii)(B) for this parameter.

15

Enclosure PG&E Letter DCL-16-015

• PG&E Response The applicable TS for the flow for the MEDT is TS 5.5.2 "Primary Coolant Sources Outside Containment." As stated in Section 2.3, "Planned Procedural Updates," in the technical report "Implementation of Alternate Source Terms Summary of Dose Analyses and Results," Revision 1, the ESF system leak testing procedures (that are part of the Boundary Leakage Program invoked by TS 5.5.2) will be updated to establish administrative acceptance criteria to ensure that the total as-tested flow hard piped to the MEDT is less than 950 cc/min of ESF system leakage and 484 cc/min of nonradioactive fluid.

NRC ARCB-RAI-11 In Enclosure Attachment 4, Section 7.2.3.1, "Containment PressureNacuum Relief Line Release," it states:

It is conservatively assumed that 40% of release flashes and is instantaneously and homogeneously mixed in the containment atmosphere, and that the activity associated with the volatiles, i.e., 100% of the noble gases and 40% of the iodine in the reactor coolant, is available for release to the environment via this pathway.

RG 1. 183 Appendix A, Regulatory Position 3. 8 states that the purge release evaluation should assume that 100% of the radionuclide inventory in the reactor coolant system liquid is released to the containment at the initiation of the LOCA and that this inventory should be based on the TS reactor coolant system equilibrium activity. RG 1. 183 Appendix A, Regulatory Position 3. 8 does not make any statements about a reactor coolant system liquid flashing fraction. However, it is the NRC staff's position that 100%

of the radionuclide inventory in the reactor coolant system liquid is released to the containment and therefore, the flashing fraction would be 100% which is conservative and meets the intent of RG 1. 183 Appendix A, Regulatory Position 3. 8.

Please provide a containment pressure I vacuum relief line release analysis that is consistent with RG 1. 183 Appendix A, Regulatory Position 3. 8, as described above or explain how the flashing fraction of 40% is more conservative than a flashing fraction of

.100%.

PG&E Response As noted in Section 7.2.3.1 of Attachment 4 of AST LAR 15-03, and in accordance with RG 1.183, Revision 0, Appendix A, Regulatory Position 3.8, the LOCA dose consequence analysis assumes that 100 percent of the radionuclide inventory in the primary coolant (assumed to be at TS levels), is released to the containment at T= 0 hours0 days <br />0 hours <br />0 weeks <br />0 months <br />.

16

Enclosure PG&E Letter DCL-16-015 As noted by the Staff, Regulatory Position 3.8 does not make any statements about a reactor coolant system liquid flashing fraction. Since no specific guidance was provided in RG 1.183, for purposes of establishing dose consequences, PG&E utilizes a flash fraction commensurate with reactor coolant conditions at full power. The calculated flash fraction is approximately 38 percent. The dose consequence analysis supporting AST LAR 15-03 uses a flash fraction of 40 percent, which is conservative when compared to the calculated value.

NRC ARCB-RAI-12 In Enclosure Attachment 4, Section 7.1, ({Control Room Design I Operation I Transport Model," it describes the basic operation of the control room ventilation system (CRVS),

provides the Mode 4 parameter values assumed in the dose consequence analysis and references a December 2012 control room tracer gas test for the maximum unfiltered in leakage to the control room envelope (CRE). However, the Mode 4 parameter values do not seem to match the description of the CRVS operation provided in Section 7. 1.

One example is that the discussion states that in Mode 4 the pressurization flow at either intake is between 650 - 900 cubic feet per minute (cfm) and the air is taken from the less contaminated intake, however, in the Mode 4 parameter table it shows the pressurization flow of 650 - 900 cfm and also includes a filtered intake of 550 - BOO cfm. It is not clear to the NRC staff if this is another outside air intake or if this is part of the recirculation flow path, therefore, please provide a simplified diagram of the CRVS and explain in further detail the operation CRVS including the specific flow rates through the components as compared to the values assumed in the dose consequence analyses.

In addition, please provide a summary describing the December 2012 CRE test results and the test configurations.

PG&E Response During CRVS Mode 4, there is no outside air intake other than the pressurization flow and unfiltered CR in leakage. As stated in Section 7.1 of Attachment 4 of AST LAR 15-03, the Mode 4 pressurization flow at either intake is between 650- 900 cfm of which 100 cfm enters the CR unfiltered due to backdraft damper leakage. Thus the filtered portion of the pressurization flow is 550- 800 cfm.

Provided below is a simplified diagram of the CRVS when in Mode 4 operation. The CRVS Mode 4 parameter values assumed in the dose consequence analyses (provided in Section 7.1 of Attachment 4 of AST LAR 15-03) are also provided in the table below.

To provide a better understanding of the CRVS operation when in Mode 4, the referenced tabular data has been expanded to include the flow symbols used in the diagram below. (Note the location numbers are discussed below.)

17

Enclosure PG&E Letter DCL-16-015 DCPP CRVS Mode 4 Operation PF + UFI UFI BDD 7 PF-BDD PF -+---...J....__---i::: >--.

Control Room FR CRVS Mode 4 Min Flow Max Flow Location(Note 1) Mode 4 CR Parameters Symbol ( cfm} (Note 2) (cfm}

1 Pressurization Flow PF 650 900 2 Backdraft damper Leakage BOD 100 100 3 Filtered Intake PF-BDD 550 800 4 CF= FR + PF-1800 2200 Charcoal Filter Flow BOD 5 Filtered Recirc Flow FR 1250 1400 6 Unfiltered lnleakage UFI 70 70 7 CR Exhaust Flow (EF) PF + UFI 720 970 Note 1: Locations 1-5 are listed on the attached operations valve identification drawings (OVIDs) presented in Figures 17 and 18. Locations 6 & 7 are CR inleakage and exhaust flow (not CRVS flows).

Note 2: For a minimum flow example, Location 4 (1800) = location 1 - location 2

+location 5 (650-100+1250).

December 2012 CRE Test Results and Configurations In December 2012, CR Habitability tracer gas in leakage testing was performed at DCPP in accordance with DCPP Procedure STP M-57, "Control Room Ventilation System Tracer Gas Test." The flow values in the above table encompass the test results.

Location numbers in the table correspond to numbers on Attachment 1, Figures 17 and 18 (DCPP operating valve identification drawings (OVID's)) and identify the approximate locations of the parameters on the OVIDs. The OVIDs are also highlighted to identify 18

Enclosure PG&E Letter DCL-16-015 the minimum in-service CRVS equipment. This assumes a safety injection signal generated from Unit 1, which causes a train of CR ventilation to swap to Mode 4 on the Unit 1 side and a pressurization fan on Unit 2 side to start.

The amount of unfiltered in leakage into the pressurized CRE was determined using the constant injection method for each of the Fans (S99 and S96) supplying pressurization air. Based on the test results of each pressurization fan, two additional tests were performed. Testing was then performed with the S99 Fan and the TSC ventilation system in operation. The final test had the S99 Fan in operation, Unit 1 in Mode 3 and Unit 2 in Mode 4. The selected alignments were based on previous tests performed at DCPP.

A total of 4 tests were performed at DCPP over the course of 4 nights to determine the total in leakage into the CRE under various system operational modes. Summaries of these test results are shown in the table below.

Summary of Results of CR In leakage for the Tested Configurations lnleakage with Tracer Gas Phase. Train Configuration Outside Air (SCFM) 1 S99, Unit 2 Mode 4 Unit 1 Idle 32 +/- 5 2 S96, Unit 1 Mode 4 Unit 2 Idle 25 +/- 10 3 S99, Unit 2 Mode 4 Unit 1 Idle + TSC 23 +/- 7 4 S99, Unit 2 Mode 4 Unit 1 Mode 3 7+/-9 NRC ARCB-RAI-13 In Enclosure Attachment 4, Section 7.2. 7, "Technical Support Center Dose," it describes the basic operation of the TSC ventilation system and then states that it utilizes the CRVS. However, it is not clear to the NRC staff how the TSC ventilation system and the CRVS are combined in operation. Therefore, please provide a simplified diagram of the TSC ventilation system/CRVS and explain in further detail the operation TSC/CRVS including the specific flow rates through the components as compared to the values assumed in the dose consequence analyses.

PG&E Response Provided below are simplified diagrams of the TSC ventilation system during Mode 1 and Mode 4 operation. To provide a better understanding of the TSC ventilation system, tabular data that present the ventilation flows (and include the flow symbols used in the diagram), are also presented.

• 19

Enclosure PG&E Letter DCL-16-015 DCPP TSC - Mode 1 Operation

~~

EF FNI

.... UFI Mode 1 DCPP TSC - Mode 4 Operation EF UFI PF

--+------+~ ~ ~ ~ ~_____.

FR Mode4 Parameter S~mbol TSC Vent S~stem Value Filtered (HEPA only) Mode 1 Intake Rate FNI 500 cfm Filtered (Carbon + HEPA) Mode 4 Intake PF 500 cfm Rate Mode 4 Filtered (Carbon + HEPA)

FR 500 cfm Recirculation Rate Unfiltered lnleakage UFI 60 cfm Mode 1: FNI + UFI Environmental Exhaust Rate EF Mode 4: PF + UFI 20

Enclosure PG&E Letter DCL-16-015 Simplified diagrams of the TSC ventilation system in combination with the CRVS during Mode 4 operation for the period 0 to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> after a LOCA, and 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> to 30 days after the LOCA, are provided in Figures 19 and 20, respectively.

Figures 19 and 20 show that the Units 1 and 2 CR emergency pressurization air intakes also serve the TSC. Thus, the redundant PG&E Design Class I radiation monitors associated with CRVS Mode 4 operation and located at each pressurization air intake (which serve to automatically select the least contaminated inlet at the beginning of the accident, and manually select the least contaminated inlet during the course of the accident), also serve the TSC.

NRC ARCB-RAI-14 In the Enclosure Section 2. 1, "Proposed Changes to Current Licensing Basis," item 22 PG&E requests the following:

Credit the following existing administrative controls reflected in plant procedures.

These administrative controls ensure the [fuel handling building] FHB is maintained at a negative pressure relative to atmosphere during movement of irradiated fuel in the spent fuel pool, thus ensuring that the environmental releases occur via the Unit vent.

• The movable wall is in place and secured.

• No exit door from the FHB is propped open.

• At least one FHB ventilation system exhaust fan is running.

Attachment 4, Section 7. 3 presents the [fuel handling accident] FHA. Credit for the above administrative controls is taken to facilitate that the post-accident environmental release of radioactivity occurs via the plant vent.

In Enclosure Attachment 4, Section 7.3, it states that the radioactivity release pathways following a FHA in the FHB are established taking into consideration the following administrative controls:

• The movable wall is put in place and secured.

• No exit door is propped open.

• One fuel handling building ventilation system (FHBVS) exhaust fan is operating.

It further states that operation of the FHBVS with a minimum of one exhaust fan operating and all significant openings administratively closed will ensure negative pressure in the FHB which will result in post-accident environmental release of radioactivity occurring via the plant vent. However, TS 3. 7.13, "FHB VS," does not require one exhaust fan to be operating, the movable wall to be in place and secured, or require the exit to be closed.

21

Enclosure PG&E Letter DCL-16-015 10 CFR 50.36(c)(2)(ii) states:

A technical specification limiting condition for operation of a nuclear reactor must be established for each item meeting one or more of the following criteria:

(A) Criterion 1 ...

(B) Criterion 2. A process variable, design feature, or operating restriction that is an initial condition of a design basis accident or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

(C) Criterion 3 .. .

(D) Criterion 4 .. .

In addition RG 1. 183, Regulatory Position 5. 1. 2 states That credit may be taken for accident mitigation features that are classified as safety-related, are required to be operable by technical specifications ... "

The administrative controls discussed above are initial conditions of the design basis FHA in the FHB radiological consequence that assumes the failure of a fission product barrier. Please explain how DCPP Technical Specification 3. 7. 13 meets 10 CFR 50.36(c)(2)(ii)(B) for these administrative control conditions and meets RG 1. 183, Regulatory Position 5. 1. 2.

PG&E Response The administrative controls to ensure that the environmental releases occur via the unit vent are not required to be in TS 3. 7.13 because they do not fall within the criteria for mandatory inclusion in the TS in 10 CFR 50.36(c)(2)(ii). They are not needed to obviate the possibility that an abnormal situation or event will give rise to an immediate threat to the public health and safety. The NRC has previously approved relocation of similar administrative controls from the TS to PG&E-controlled procedures. The NRC approved 14 relocated TS from the TS to the PG&E ECGs and FSAR and associated PG&E-controlled procedures as part of the conversion to the improved TS in NRC License Amendments 135 for each unit dated May 28, 1999. The relocated TS and associated ECGs and FSARs were:

Relocated TS Title ECG/FSAR 3/4.1.2.1 Reactivity Control Systems - Boration ECG 8.5 Systems- Flow Path- Shutdown 3/4.1.2.3 Reactivity Control Systems - Charging ECG 8.6 Pump - Shutdown 3/4.1.2.4 Reactivity Control Systems - Charging ECG 8.7 Pumps - Operating 22

Enclosure PG&E Letter DCL-16-015 Relocated TS Title ECG/FSAR 3/4.1.2.5 Reactivity Control Systems - Borated ECG 8.8 Water Source - Shutdown 3/4.1.2.6 Reactivity Control Systems- Borated ECG 8.9 Water Sources - Operating 3/4.3.3.2 Movable lncore Detectors ECG 48.1 3/4.3.3.4 Meteorological Instrumentation ECG 40.1 3/4.9.3 Refueling Operations - Decay Time ECG 42.1 3/4.9.5 Refueling Operations - Communications ECG 42.2 3/4.9.6 Refueling Operations - Manipulator ECG 42.3 Crane and Auxiliary Hoist 3/4.9.7 Refueling Operations - Crane Travel - ECG 42.4 Fuel Handling Building 3/4.9.10.2 Refueling Operations -Water Level - ECG 42.5 Reactor Vessel - (Control Rods) 3/4.9.13 Spent Fuel Shipping Cask Movement FSAR 9.1 3/4.10.4 Special Test Exceptions - Position ECG 41.2 Indication System - Shutdown The NRC provided the following conclusion for the 14 relocated specifications:

The 14 relocated specifications from the CTS discussed above are not required to be in the ITS because they do not fall within the criteria for mandatory inclusion in the TS in 10 CFR 50.36(c)(2)(ii). They are not needed to obviate the possibility that an abnormal situation or event will give rise to an immediate threat to the public health and safety. In addition, the NRC staff finds that sufficient regulatory controls exist under the regulations cited above to maintain the effect of the provisions in these specifications. The NRC staff has concluded that appropriate controls have been established for all of the current specifications, information, and requirements that are being moved to an EGG and the FSAR.

The administrative controls to ensure the FHB is maintained at a negative pressure relative to atmosphere during movement of irradiated fuel in the spent fuel pool (SFP) will be included in the UFSAR as shown in Section 15.5.22.2.2, "Activity Release Transport Model," of Attachment 4 to the Enclosure of PG&E Letter DCL-15-152, dated December 17, 2015, which states in part:

The radioactivity release pathways following an FHA in the FHB are established taking into consideration the following Administration Controls:

23

Enclosure PG&E Letter DCL-16-015 During fuel movement in the FHB:

a. The movable wall is put in place and secured

b. No exit door is propped open

c. One FHBVS exhaust fan is operating (The supply fan flow (if operating) has been confirmed by design to have less flow than the exhaust fan)

These administrative controls are already included in existing Diablo Canyon Operating Procedure OP B-8D, "Refueling Prerequisites."

NRC ARCB-RAI-15 In Enclosure Attachment 4, Section 7.3 it has been determined that for the FHA in the FHB, the actual release rate lambda based on the FHBVS exhaust (i.e., 8. 7 hr-1) is larger than the release rate applicable to {(a 2-hr release" per regulatory guidance (i.e.,

3.45 hr-1). Thus the larger exhaust rate lambda associated with FHBVS operation plus the exhaust rate lambda for the 500 cfm outleakage is utilized in the analysis.

Please provide the technical basis for why this approach is conservative for control room doses.

PG&E Response Although both release rate lambdas (8.7 h( 1 and 3.45 h( 1) will release essentially all the airborne activity in the FHB in two hours, the release case with a larger release rate lambda will release activity faster than that with a smaller release rate lambda.

At the beginning of the accident (t = 0), the activity release rate (Ci/hr) for the 8.7 h( 1 release lambda case is 2.5 times greater than that for the 3.45 h( 1 release lambda case. This multiplier decreases as release goes on . At t = 10.6 min (0.176 hr, ln(8.7/3.45)/(8.7-3.45) = 0.176), the activity release rates for the two cases are the same. After 10.6 minutes, the activity release rate for the 8.7 h( 1 release lambda case is less than that for the 3.45 h( 1 release lambda case.

If the CR intake remained constant over the 2-hour release period, the resulting doses for the two cases would be the same. However, the DCPP CR intake switches from Mode 1 normal operation (4200 cfm plus or minus 10 percent unfiltered intake), to Mode 4 pressurization operation (550 cfm filtered intake, 70 cfm unfiltered inleakage, and 100 cfm unfiltered leakage via the backdraft dampers) after detection of contaminated air by the CR intake monitors. Thus a greater activity release rate at the earlier stage of the accident will result in a greater activity intake into the CR and consequently more conservative CR doses. A more detailed quantitative explanation is provided below:

24

Enclosure PG&E Letter DCL-16-015 Use of a release rate lambda of 8.7 h( 1 (vs. 3.45 h( 1) based on the FHB ventilation exhaust fans to calculate the CR doses following a FHA in the FHB impacts:

• the monitor response time,

• the activity intake into the CR, and

• the CR dose contribution due to the 500 cfm outleakage from the doors/openings of the fuel handling building.

As noted in LAR Enclosure, Attachment 4, Section 7.3, the instrument time constant for the DCPP CR Intake Monitors (1-RM-25, 1-RM-26, 2-RM-25, and 2-RM-26) is 2 seconds. The monitor response time for a design basis FHA (which will release a significant amount of activity to the atmosphere), is less than 1 second for both release rate lambdas (0.155 sec for the 8. 7 h( 1 release lambda, 0.42 sec for the 3.45 h(1 release lambda, based on equation presented in Attachment 4, Section 7.3). The analysis conservatively assumes a 20 second monitor response time. The delay time for switching to Mode 4 operation is 32 seconds, which includes the signal processing time (2 seconds) and the damper closure time (1 0 seconds).

• As explained earlier, at the beginning of the accident, the activity release rate for the 8.7 h( 1 release lambda case is 2.5 times greater than that for the 3.45 h( 1 release lambda case. Therefore, the CR activity intake during the 32 second Mode 1 operation is approximately 2.5 times greater for the 8. 7 h( 1 release lambda case. At t = 32 sec, the CR ventilation operation is switched to Mode 4 which results in less contaminated air entering the CR (i.e., due to the location of the pressurization intakes and associated smaller X/Qs, and a filter efficiency of 93 percent for elemental/organic iodine). The analysis also considers the dose contribution from (a) unfiltered inleakage, and (b) the pressurization intake backdraft damper bypass leakage, into the CR. From t = 32 sec to t = 10.6 min, the CR activity intake rate for the 8.7 hr1 lambda case remains greater than that for the 3.45 h( 1 lambda case. After 10.6 minutes, the CR activity intake rate for the

8. 7 h( 1 release lambda case is less than that for the 3.45 h( 1 release lambda case.

Because the atmospheric dispersion factors at the pressurization intakes are substantially smaller and the pressurization flow is filtered, the CR dose due to releases from the plant vent is dominated by a) the activity intake during the initial 32 second Mode 1 operation (4200 cfm plus or minus 10 percent unfiltered intake) and b) the 70 cfm unfiltered inleakage during the Mode 4 operation. The above two pathways result in 98.6 percent of the total dose from the plant vent release for the 8.7 h( 1 release lambda case.

Since the average X/Q value at the Unit 1 and 2 normal operation intakes and the X/Q value for unfiltered in leakage location are comparable, the ratio of activity intake during the initial 32 second verses that due to 70 cfm unfiltered in leakage is approximately

[4620 cfm x 8.7 h( 1 x (32/3600) hr] I [70 cfm x 1.0] = 5.1, for the 8.7 h( 1 release lambda case; and is approximately [4620 cfm x 3.45 h( 1 x (32/3600) hr] I [70 cfm x 1.0] = 2.0, for the 3.45 h( 1 release lambda case. The above comparison clearly demonstrates that 25

Enclosure PG&E Letter DCL-16-015 the activity intake during the initial 32 seconds of the accident is the most significant contributor to the total CR dose due to FHB ventilation release.

A smaller release rate lambda for the FHB ventilation release will increase slightly the CR dose contributed by the 500 cfm outleakage from the doors/openings because more activity is retained in the building and available for the outleakage release. However, the CR dose contribution from the 500 cfm outleakage (equivalent to 9.48E-2 h( 1 release rate lambda) constitutes only 3.8 percent of the total dose from both the plant vent and the FHB outleakage. This small dose contribution from the FHB outleakage is due to a release rate that is two orders of magnitude smaller than the plant vent release rate, and a X/Q value that is greater than that applicable to the plant vent release, by only a factor of 4. Thus, a small increase of this dose component will not change the conclusion that using the 8.7 h( 1 release rate lambda for the FHB ventilation release is conservative.

• Based on above, and given that the radiation monitor response time is essentially instantaneous in both cases, it is concluded that using the 8. 7 h( 1 release rate lambda for the FHB ventilation release is more conservative than using the 3.45 h( 1 release rate lambda in calculating the CR dose for a FHA in the FHB.

NRC ARCB-RAI-16 Appendix B of Regulatory Guide (RG) 1.183, Regulatory Position 1.1 states:

The number of fuel rods damaged during the accident should be based on a conservative analysis that considers the most limiting case. This analysis should consider parameters such as the weight of the dropped heavy load or the weight of a dropped fuel assembly (plus any attached handling grapples), the height of the drop, and the compression, torsion, and shear stresses on the irradiated fuel rods. Damage to adjacent fuel assemblies, if applicable (e.g., events over the reactor vessel), should be considered.

In Enclosure Attachment 4 Section 7. 3, it postulates that a spent fuel assembly is dropped during refueling in the spent fuel pool (SFP) located in the FHB, or in the reactor cavity located in containment and that all the rods in the dropped assembly are damaged.

The NRC staff reviewed the licensee's FHA analysis. The FHA activity is assumed to be released from (1) the damaged fuel via the SFP to the FHB, or (2) from the damaged fuel via the reactor cavity to the containment, from which it is assumed to be released to the environment within a two hour period as an unfiltered release via the plant vent or the containment equipment hatch, respectively. Credit was taken for the operation of one fuel handling building ventilation system exhaust fan, and the closure of the FHB exit doors. Credit was not taken for the fuel handling building ventilation system filtration. When evaluating the dose to operators in the control room, it was assumed 26

Enclosure PG&E Letter DCL-16-015 that the control room normal intake radiation monitors would initiate CRVS pressurization mode 4 at 22 seconds following the start of the event.

DCPP technical specification (TS) 3.3. 7, "Control Room Ventilation System (CRVS)

Actuation Instrumentation," states that the CRVS actuation instrumentation for each function in Table 3.3. 7-1 shall be operable, and is applicable according to Table 3.3. 7-1, which is during MODES 1, 2, 3, 4, 5, and 6, and during movement of recently irradiated fuel assemblies.

DCPP TS 3.3.8, "Fuel Handling Building Ventilation System (FHBVS) Actuation Instrumentation," states that the FHBVS actuation instrumentation for each function in Table 3. 3. 8-1 shall be operable, and is applicable according to Table 3. 3. 8-1, which is during movement of recently irradiated fuel assemblies in the fuel handling building.

DCPP TS 3. 7. 10, "Control Room Ventilation System (CRVS)," states that two CRVS trains shall be operable, and is applicable during MODES 1, 2, 3, 4, 5, and 6, and during movement of recently irradiated fuel assemblies.

DCPP TS 3.7.13, "Fuel Handling Building Ventilation System (FHBVS)," states that two FHBVS trains shall be operable and is applicable during movement of recently irradiated fuel assemblies in the fuel handling building.

DCPP TS 3. 7. 15, "Spent Fuel Pool Water Level," states that the spent fuel pool water level shall be ;::: 23 feet over the top of irradiated fuel assemblies seated in the storage racks and is applicable during movement of irradiated fuel assemblies in the spent fuel pool.

DCPP TS 3. 9. 7, "Refueling Cavity Water Level," states that the refueling cavity water level shall be maintained ;::: 23 feet above the top of reactor vessel flange and is applicable during movement of irradiate fuel assemblies within containment.

Given that the licensee voluntarily has requested a change to its licensing basis, please explain how the proposed revised fuel handling accident analysis meets or bounds Regulatory Guide 1. 183, Appendix B Regulatory Position 1. 1. Specifically, taking into account that the technical specification applicability does not:

1. Require CRVS actuation instrumentation or CRVS to be operable when in no mode during the movement of other radioactive loads (those other than recently irradiated fuel such as sources) in the SFP and other loads (such as fresh fuel) over irradiated fuel in the FHB or containment.

2. Require the FHBVS actuation instrumentation to be operable, a FHBVS exhaust fan to be running, or exit doors to be closed, or any of the equipment to be operable during movement of other radioactive loads (not defined as "recently" irradiated fuel) or loads (such as fresh fuel assemblies or sources, etc.) over irradiated fuel in the FHB.

27

Enclosure PG&E Letter DCL-16-015

3. Require the SFP or reactor cavity water level to be at least 23 feet during the movement of other radioactive loads (those other than irradiated fuel such as sources) and other loads (such as a fresh fuel) over irradiated fuel in the FHB or containment.

In addition, clarify how the revised analysis determines the most limiting case and how the fuel handling analysis shows that the limiting case is not the drop of a fuel assembly or object other than a recently irradiated fuel assembly.

PG&E Response The AST FHA accident analysis assumes the movement of an irradiated fuel assembly at the earliest permitted time of greater than or equal to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> from removal of the critical reactor core. The FHA assumption that the failure of all rods in one entire fuel assembly that was irradiated is consistent with the current limiting licensing basis limiting case that was approved by the NRC in License Amendments 8 (Unit 1) and 6 (Unit 2), dated May 30, 1986. The use of the AST methodology does not result in a change in the limiting case of one entire fuel assembly that was irradiated. In the safety evaluation for Amendments 8 and 6, the NRC concluded:

Regardless of how many spent fuel assemblies are stored in the pool, all fuel rods in one entire, freshly discharged fuel assembly were assumed to be breached, releasing their entire gap activity in accordance with conservative assumptions in Regulatory Guide 1.25. The assumption that all fuel rods in one assembly rupture is conservative because the kinetic energy available for causing damage to a fuel assembly dropped through water is fixed by the drop distance. The kinetic energy associated with the maximum drop height for a fuel handling accident is not considered sufficient to rupture the equivalent number of fuel rods of one assembly in both the dropped assembly and the impacted assembly. The accident evaluation conservatively assumes all rods in one entire assembly rupture and release their gap activity.

The movement of loads other than irradiated fuel, including control rods, is controlled through PG&E administrative controls to ensure that the equipment used to handle loads within the reactor cavity and SFP regions will function as designed and that the equipment has sufficient load capacity for handling loads. These administrative controls used to be contained in the TS; however, the NRC previously approved relocation of these administrative controls from the TS to PG&E-controlled procedures as part of the conversion to the improved TS in NRC License Amendments 135 for each unit dated May28, 1999. The NRC approval was based on the administrative requirements not falling within the criteria for mandatory inclusion in the TS in 10 CFR 50.36(c)(2)(ii).

Originally, the TS contained TS 3/4.9.6, "Manipulator Crane and Auxiliary Hoist," that contained requirements for (1) the manipulator crane, to ensure a minimum capacity of 3250 pounds and an overload cutoff limit of no more than 2700 pounds, and (2) 28

Enclosure PG&E Letter DCL-16-015 the auxiliary hoist, to ensure a minimum capacity of 700 pounds and a load indicator which is used to prevent lifting loads greater than 600 pounds. The NRC approved relocation of these original TS requirements to PG&E controlled ECGs as part of Amendments 135, and PG&E included the requirements in ECG 42.3, "Refueling Operations - Manipulator Crane." The NRC provided the following conclusion for the relocated TS 3/4.9.6:

Administrative controls exist to ensure that the equipment used to handle fuel within the reactor pressure vessel will function as designed and that the equipment has sufficient load capacity for handling fuel assemblies or control rods or both.

The refueling machine is designed with interlocks to provide overload limits to prevent damage to refueling equipment and fuel assemblies. These limits are not taken credit for to mitigate the consequences of a DBA; nor do these limits represent initial condition assumptions of an accident analysis. Although these limits represent operating restrictions and Criterion 2 of the Final Policy Statement and 10 CFR 50.36 includes operating restrictions, Criterion 2 applies only to those operating restrictions required to preclude analyzed accidents and transients.

Therefore, the Final Policy Statement and 10 CFR 50.36 criteria for including these requirements in the ITS are not met. An EGG is an acceptable licensee-controlled document for this information. This relocation is acceptable.

Originally, the TS contained TS 3/4.9.7, "Crane Travel* Spent Fuel Storage Areas,"

that contained requirements for the SFP crane that required that loads in excess of 2500 pounds not travel over fuel assemblies in the SFP (1) to limit the effect of a dropped load on the release of the radioactivity from a single irradiated fuel assembly (i.e., limit the effect to the gap activity), and (2) to assure that a dropped load does not distort the fuel in the SFP racks so as to cause criticality. The NRC approved relocation of these original TS requirements to PG&E controlled ECGs as part of Amendments 135, and PG&E included the requirements in ECG 42.4, "Refueling Operations - Crane Travel - Fuel Handling Building." The NRC provided the following conclusion for the relocated TS 3/4.9.7:

The administrative monitoring of loads moving over the fuel storage racks serves as a backup to the crane interlocks and physical stops. Although the specification supporls the maximum refueling accident assumption in the DBA, the crane travel limits are not monitored and controlled during operation; they are checked on a periodic basis to ensure their operability. Although this limit represents operating restrictions and Criterion 2 of the Final Policy Statement and 10 CFR 50.36 includes operating restrictions, Criterion 2 applies only to those operating restrictions required to preclude analyzed accidents and transients. This limit is not required to preclude analyzed accidents and transients. Therefore, the Final Policy Statement and 10 CFR 50.36 criteria for including these requirements in the ITS are not met. An EGG is an acceptable licensee-controlled document for this information. This relocation is acceptable.

29

Enclosure PG&E Letter DCL-16-015 Originally, the TS contained TS 3/4.9.1 0.2, "Reactor Vessel Water Level," that contained requirements on the reactor vessel water level during movement of control rods within the reactor vessel in Mode 6. The NRC approved relocation of these original TS requirements to PG&E controlled EGGs as part of Amendments 135, and PG&E included the requirements in ECG 42.5, "Refueling Operations -Water Level -

Reactor Vessel (Control Rods)." The NRC provided the following conclusion for the relocated TS 3/4.9.10.2:

The specification limits the minimum water over the irradiated fuel in the reactor core during the movement of controls rods within the reactor vessel while in Mode 6. Although this minimum water /eve/limit is being relocated to an EGG, there is the remaining requirement in ITS 3. 9. 7 (CTS 314. 9. 10. 1) that limits the minimum water level above the reactor core during movement of irradiated fuel assemblies within the containment. The limits in CTS 314. 9. 10.2 are not an initial condition of a DBA or transient. Although this limit represents operating restrictions and Criterion 2 of the Final Policy Statement and 10 CFR 50.36 includes operating restrictions, Criterion 2 applies only to those operating restrictions required to preclude analyzed accidents and transients. Therefore, the Final Policy Statement and 10 CFR 50.36 criteria for including these requirements in the ITS are not met.

An EGG is an acceptable licensee-controlled document for this information. This relocation is acceptable.

To account for the possibility that a fuel assembly that has decayed such that the CRVS will not actuate, an additional analysis case was performed for determining the dose to the CR operator for a FHA in the FHB and a FHA in Containment. The case assumed a delayed FHA at fuel offload or a FHA during reload occurs at a time when the fuel has decayed to such an extent that the radiation environment at the CR normal intake radiation monitors is just below the CRVS setpoint (CRVS Mode 4 is not initiated). This analysis case is discussed in PG&E Letter DCL-15-152, dated December 17, 2015, Enclosure Attachment 2, Section 7.3, "Fuel Handling Accident (FHA)." The analyses determined that the dose consequences of a DBA FHA (72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> of decay after reactor shutdown) bound those associated with the delayed FHA for both the FHA in the FHB and the FHA in the containment.

30

Enclosure Attachment 1 PG&E Letter DCL-16-015 Figure 1 General Plant Layout/Areas showing Control Room and TSC Location relative to U1/U2 Containments j I I 1. I I ~ i !r 1

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Enclosure Attachment 1 PG&E Letter DCL-16-015 Figure 3 Elevation View of the Containment Wall and Control Room

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Enclosure Attachment 1 PG&E Letter DCL-16-015 Figure 5 Cloud Shine through CR Doorway at Column L t---8.72'-----'

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Enclosure Attachment 1 PG&E Letter DCL-16-015 Figure 6 Elevation View of MER, CR, CRVS Filters and Penetrations 0 I I

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Enclosure Attachment 1 PG&E Letter DCL-16-015 Figure 7 Penetrations along Column Line L (i.e., the wall separating Main CR and Mechanical Room) 18 I

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Penetrations after Col 18 are mirror image to those shown, noting that only one CRVS filter train is assumed to be operable.

MER- Part of Pressure boundary- no cloud source X

Note: The Centerline of each Penetration is located 17'-3" above the Control Room Floor 3

Enclosure Attachment 1 PG&E Letter DCL-16-015 Figure 8 Containment and Control Room Model Plan View 121 I G -+ ---~*

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CR- floor 2' (cone.) : Ceiling 3'-4" (cone.)

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Enclosure Attachment 1 PG&E Letter DCL-16-015 Figure 9 Containment SW-QADCGGP Model SPH 1 & 2

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Enclosure Attachment 1 PG&E Letter DCL-16-015 Figure 12 Gamma-Ray Trace Model for RWST Shine I .4~

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Enclosure Attachment 1 PG&E Letter DCL-16-015 Figure 13 Technical Support Center Plan View C.-

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Enclosure Attachment 1 PG&E Letter DCL-16-015 Figure 14a- Labyrinths in East TSC wall Door BU 201-2 I

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Enclosure Attachment 1 PG&E Letter DCL-16-015 Figure 14c - Labyrinths in West TSC wall Door BU 206

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Enclosure Attachment 1 PG&E Letter DCL-16-015 Figure 14d- Labyrinths in West TSC wall Door BU 208 Records Management Area Gypsum Wall

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Enclosure Attachment 1 PG&E Letter DCL-16-015 Figure 15 Mechanicai/HVAC Room adjacent to the NRC Office TSC- Mechanical/ HVAC Room Filter inlet ~

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Enclosure Attachment 1 PG&E Letter DCL-16-015 Figure 16 TSC Filter Scatter analysis Dose map of Mechanical Room and NRC Office plotted from the MCNPS mesh tally (mrem) 43 . 9 --t

---

.62

Enclosure Attachment 1 PG&E Letter DCL-16-015 FIGURE 17- CRVS Operating Valve Identification Drawings (OVID's) 161 62 2

Enclosure Attachment 1 PG&E Letter DCL-16-015 FIGURE 18- CRVS Operating Valve Identification Drawings (OVID's) 67

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Enclosure Attachment 1 PG&E Letter DCL-16-015 Figure 19 TSC/CR HVAC Alignment (TSC- 0-2 hrs Post-LOCA; CR- 44 sec- 2 hrs Post-LOCA)

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CR Mode 4 by automatiic initiation (44 seconds to 30 days after LOCA)

Enclosure Attachment 1 PG&E Letter DCL-16-015 Figure 20 TSC/CR HVAC Alignment 2 hrs - 30 days Post-LOCA

~ Infiltration Ill"'

TSC TSC Mode 4 by manual operation (2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> after LOCA continuing to 30 days)

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